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BWRVIP-319NP: BWR Vessel and Internals Project, Testing and Evaluation of the Hatch Unit 2 120 Degrees Surveillance Capsule
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t=~1211 ELECTRIC POWER ti=I-RESEARCH INSTITUTE 2018 TECHNICAL REPORT BWRVIP-319NP: BWR Vessel and Internals Project Testing and Evaluation of the Hatch Unit 2 120° Surveillance Capsule

BWRVIP-319NP: BWR Vessel and Internals Project Testing and Evaluation of the Hatch Unit 2 120° SurveJllance Capsule 3002013103 Final Report, August 2018 EPRI Project Manager N. Palm All or a portion of the requirements of the EPRI Nuclear Quality Assurance Program apply to this product.

'NO ELECTRIC POWER RESEARCH INSTITUTE 3420 Hillview Avenue, Palo Alto, California 94304-1338

PO Box 10412, Palo Alto, California 94303-0813

USA 800.313.3774

650.855.2121

askepri@epri.corn

www.epri.corn

DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THIS DOCUMENT WAS PREPARED BY THE ORGANIZATION(S) NAMED BELOW AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCH INSTITUTE, INC. (EPRI). NEITHER EPRI, ANY MEMBER OF EPRI, ANY COSPONSOR, THE ORGANIZATION(S)

BELOW, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM:

(A) MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED, (I)

WITH RESPECT TO THE USE OF ANY INFORMATION, APPARATUS, METHOD; PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT, INCLUDING MERCHANTABILITY AND FITNESS FOR A PARTICULAR PURPOSE, OR (II) THAT SUCH USE DOES NOT INFRINGE ON OR INTERFERE WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTY'S INTELLECTUAL PROPERTY, OR (Ill) THAT THIS DOCUMENT IS SUITABLE TO ANY PARTICULAR USER'S CIRCUMSTANCE; OR (B) ASSUMES RESPONSIBILITY FOR ANY DAMAGES OR OTHER LIABILITY WHATSOEVER (INCLUDING ANY CONSEQUENTIAL DAMAGES, EVEN IF EPRI OR ANY EPRI REPRESENTATIVE HAS BEEN ADVISED OF THE POSSIBILITY OF SUCH DAMAGES) RESULTING FROM YOUR SELECTION OR USE OF THIS DOCUMENT OR ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT.

REFERENCE HEREIN TO ANY SPECIFIC COMMERCIAL PRODUCT, PROCESS, OR SERVICE BY ITS TRADE NAME, TRADEMARK, MANUFACTURER, OR OTHERWISE, DOES NOT NECESSARILY CONSTITUTE OR IMPLY ITS ENDORSEMENT, RECOMMENDATION, OR FAVORING BY EPRI.

THE FOLLOWING ORGANIZATIONS PREPARED THIS REPORT:

Electric Power Research Institute (EPRI)

MP Machinery & Testing, LLC TransWare Enterprises Inc.

THE TECHNICAL CONTENTS OF THIS PRODUCT WERE PREPARED IN ACCORDANCE WITH THE EPRI QUALITY PROGRAM MANUAL THAT FULFILLS THE REQUIREMENTS OF 10 CFR 50 APPENDIX B. THIS PRODUCT IS SUBJECT TO THE REQUIREMENTS OF 10 CFR PART 21.

CERTIFICATION OF CONFORMANCE CAN BE OBTAINED FROM EPRI.

CONTRACTUAL ARRANGEMENTS BETWEEN THE CUSTOMER AND EPRI MUST BE ESTABLISHED BEFORE QUALITY APPLICATION TO,ASSURE FULFILLMENT OF QUALITY

. PROGRAM REQUIREMENTS.

NOTE For further information about EPRI, call the EPRI Customer Assistance Center at 800.313.3774 or e-mail askepri@epri.com.

Electric Power Research Institute, EPRI, and TOGETHER... SHAPING THE FUTURE OF ELECTRICITY are registered service marks of the Electric Power Research Institute, Inc.

ACKNOWLEDGMENTS The following organizations prepared this report:

Electric Power Research Institute (EPRI)

Principal Investigators N. Palm E.Long MP Machinery & Testing, LLC 2161 Sandy Drive State College, PA 16803 Principal Investigator Dr. M. P. Manahan, Sr.

Trans Ware Enterprises Inc.

1565 Mediterranean Drive Sycamore, IL 60178 Principal Investigators D. Jones K. Watkins Thi~ report describes research sponsored by EPRI and its BWRVIP participating members.

This publication is a corporate document that should be cited in the literature in the following manner:

BWRVIP-319NP: BWR Vessel and Internals Project: Testing and Evaluation of the Hatch Unit 2 120° Surveillance Capsule. EPRI, Palo Alto, CA: 2018. 3002013103.

111 i

PRODUCT DESCRIPTION In the late 1990s, EPRI's Boiling Water Reactor Vessel and Internals Project (BWRVIP)

Integrated Surveillance Program (ISP) was developed to improve the surveillance of the U.S.

BWR fleet. This report describes testing and evaluation of the Edwin I. Hatch Nuclear Power Plant (HNPP) Unit 2 120° surveillance capsule. These results will be used to monitor embrittlement as part of the BWRVIP ISP.

Background

The BWRVIP ISP represents a major enhancement to the process of monitoring embrittlement for the U.S. fleet ofBWRs_. The ISP optimizes surveillance capsule tests while at the same time maximizing the q1Jantity and quality of data, resulting in a more cost-effective program. The BWRVIP ISP provides m_ore representative data that can be used to assess embrittlement in reactor pressure vessel (RPV) beltline materials and thereby improve trend curves in the BWR range of irradiation conditions.

Objectives

. Neutron irradiation exposure reduces the toughness ofreactor vessel steel plates, welds, and forgings. The objectives of this project were twofold:

To document the results of neutron dosimetry and Charpy V-notch toughness tests for the surveillance materials (plate heat C8554 and weld heat 51912) in the Hatch Unit 2 120° surveillance capsule.

To compare the r~sults with the embrittlement trend prediction of U.S. Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.99, Revision 2.

Approach The Hatch Unit 2 120°.surveillance capsule had been irradiated in the reactor since plant startup.

The surveillance capsule contained flux wires for neutron flux monitoring, Charpy V-notch impact test specimens, and tensile specimens. The project team removed the capsule from the reactor in 2017 and transported it to facilities for testing and evaluation. The team first used dosimetry testing to gather information about the neutron fluence accrual of specimens froin the capsule. They next performed a neutron transport calculation in accordance with U.S. NRC Regulatory Guide 1.190 and compared it to results from the dosimetry testing. Testing ofCharpy V-notch specimens was performed according to the American Society for Testing and Materials (ASTM) standards.

Results and Findings The report includes capsule neutron exposure and Charpy V-notch test results for Hatch Unit 2 surveillance plate heat C8554 and surveillance weld heat 51912. The project compared irradiated Charpy data to unirradiated data in order to determine the shifts in Charpy index temperatures for V

the surveillance plate and weld materials due to irradiation. The measured shift for the surveillance plate and weld are less than the predicted shift plus margin using Regulatory Guide 1.99, Revision 2. Also covered are measurement of flux wires, determination of a fluence value for the 120° surveillance capsule, and calculation of a revised fluence value for the previously tested 30° surveillance capsule.

Applications, Value, and Use Results of this work will be used in the BWRVIP ISP that integrates individual BWR surveillance programs into a single program. The ISP provides data of high quality to monitor BWR vessel embrittlement. The !SP results in significant cost savings to the BWR fleet and provides more accurate monitoring of embrittlement in BWR vessels.

Keywords BWR Charpy V-notch testing Mechanical properties Radiation embrittlement Reactor pressure vessel integrity Reactor vessel surveillance program Vl

e ~~, 1 ELECTRIC POWER 1-1c;;;;,

RESEARCH INSTITUTE EXECUTIVE

SUMMARY

De.liverable Number: 3002013103 Product Type: Technical Report Product

Title:

BWRVIP-319NP: BWR Vessel and Internals Project: Testing and Evaluation of the Hatch Unit 2 120° Surveillance Capsule

(

PRIMARY AUDIENCE: Plant engineers responsible for reactor vessel integrity SECONDARY AUDIENCE: Boiling Water Reactor Vessel and Internals Project (BWRVIP) program owners KEY RESEARCH QUESTION The objectives of this project were To withdraw and test the Hatch Unit 2 120° surveillance capsule per the approved test matrix of the BWRVIP Integrated Surveillance Program (ISP) (BWRVIP-86, Revision 1-A; EPRI report 1025144, 2012).

To document the results of neutron dosimetry and Charpy V-notch toughness tests for the surveillance materials (plate heat C8554 and weld heat 51912) per American Society for Testing and Materials (ASTM) E185-82 and determine capsule fluence per U.S. Nuclear Regulatory Commission (U.S. NRC)

Regulatory Guide 1.190.

To compare the results with embrittlement trend predictions of U.S. NRC Regulatory Guide 1.99, Revision 2.

RESEARCH OVERVIEW The BWRVIP ISP combines individual BWR surveillance programs into a single program that monitors the reduction in toughness of reactor vessel steel plates, welds, and forgings as a result of neutron irradiation exposure. The Hatch Unit 2 120° surveillance capsule was withdrawn and tested per the schedule in BWRVIP-86, Revision 1-A. The capsule had been irradiated in the reactor since plant startup and contained flux wires for neutron flux monitoring, Charpy V-notch impact test specimens, and tensile specimens. The project team removed the capsule from the reactor in 2017 and transported it to facilities for testing and evaluation. The team first used dosimetry testing to gather information about the neutron fluence accrual of the capsule specimens. Next, the team performed a neutron transport calculation in accordance with Regulatory Guide 1.190 and compared it to the dosimetry test results. Testing of Charpy V-notch specimens was performed according to ASTM standards.

KEY FINDINGS The report includes capsule neutron exposure and Charpy V-notch test results for Hatch Unit 2 surveillance plate heat C8554 and su!:"eillance weld heat 51912.

The project team compared irradiated Charpy data to unirradiated data in order to determine the shifts in Charpy index temperatures for the surveillance plate and surveillance weld materials due to irradiation.

The measured shifts for the surveillance plate and surveillance weld are less than the predicted shift plus margin using Regulatory Guide 1.99, Revision 2.

Vll

e,=,121 ELECTRIC POWER RESEARCH INSTITUTE EXECUTIVE

SUMMARY

Researchers measured flux wires, performed a fluence calculation to determine the fluence for the 120° surveillance capsule, and updated fluence values for the previously tested 30° surveillance capsule.

WHY THIS MATTERS Results of this work will be used in the BWRVIP ISP, which is utilized by U.S. BWR fleet owners to satisfy the requirements of 10 CFR 50 Appendix G and Appendix H. The ISP provides high quality data to monitor BWR vessel embrittlement. The ISP results in significant cost savings for the BWR fleet and provides more accurate monitoring of embrittlement in BWR vessels. Plants for which the Hatch Unit 2 surveillance materials are assigned as the representative surveillance materials under the ISP must consider these test results in development of vessel integrity evaluations and plant operating limit curves.

HOW TO APPLY RESULTS Instructions for use of ISP data are provided in the following technical report:

BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014. 3002003144.

LEARNING AND ENGAGEMENT OPPORTUNITIES The program plan for the BWRVIP ISP is described in the following technical report: BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012. 1025144.

EPRI CONTACT: Nathan A. Palm, Principal Technical Leader, npalm@epri.com PROGRAM: BWRVIP IMPLEMENTATION CATEGORY: Category 1 - Regulatory Together... Shaping the Future of Electricity Electric Power Research Institute 3420 Hillview Avenue, Palo Alto, California 94304-1338

PO Box 10412, Palo Alto, California 94303-0813 USA 800.313.3774

650.855.2121

askepri@epri.com

www.epri.com

© 2018 Electric Power Research Institute (EPRI), Inc. All rights reserved. Elec.tric Power Research Institute, EPRI, and TOGETHER... SHAPING THE FUTURE OF ELECTRICITY are registered service marks of the Electric Power Research Institute, Inc.

I CONTENTS ABSTRACT................................................................................. ;.............................................. v EXECUTIVE

SUMMARY

......................*................................................................................... vii 1 INTRODUCTION................................................................................................................... 1-1 1.1 Implementation Requirements........................................................................................ 1-2 2 MATERIALS AND TEST SPECIMEN DESCRIPTION.......................................................... 2-1 2.1 *Dosimeters.................................................................................................................... 2-1 2.2 Test Materials................................................................................................................ 2-1 2.2.1 Capsule Loading Inventory :................................................................................... 2-1 2.2.2 Material Description............................................................................................... 2-6 2.2.3 Chemical Composition............................. :*********************************************************.... 2-7 2.2.4 CVN Baseline Properties.......................................................................................2-7 2.3 Capsule Opening...................................... :.................................................................. 2-18 3 NEUTRON FLUENCE CALCULATION................................................................................ 3-1 3.1 Description of the Reactor System........ :***********************************************************************3-2 3.1.1 Overview of the Reactor System Design................................................................ 3-2 3.1.2 'Reactor System Mechanical Design Inputs............................................................ 3-2 3.1.3 Reactor System Material Compositions..................................,.............................. 3-4 3.1.4 Reactor Operating Data lnputs............................................................................... 3-6 3.1.4.1 Core Configuration and Fuel Designs............ *................................................. 3-6 3.1 A.2 Reactor Power History.................................................................................... 3-6 3.1.4.3 Reactor-State-Point Data................................................................................ 3-6.

3.1.4.4 Reactor Coolant Properties............................................................................ 3-9

3.2 Methodology................................ *................................................................................ 3-1 O 3.2.1 Computational Method..................................................... :................................,.. 3-1 o 3.2.2 Fluence Model...... :...................................'.-............ :.............................................. 3-11 IX

3.2.2.1 Geometry Model................................... :....................................................... 3-14 3.2.2.2 Reactor Core and Core Reflector................................................................. 3-15 3.2.2.3 Reactor Core Shroud.................................................................................... 3-16 3.2.2.4 Downcomer Region...................................................................................... 3-17 j

3.2.2.4.1 Jet Pumps.................................................................................... ;.... :... 3-17 3.2.2.4.2 Surveillance Capsules......................... :................................................ 3-17 3.2.2.4.3 Shroud Repair Tie Rods....................................................................... 3-18 3.2.2.5 Reactor Pressure Vessel.............................................................................. 3-18 3*.2.2.6 Thermal Insulation........................................................................................ 3-18 3.2.2.7 Inner and Outer Cavity................................................................................. 3-18 3.2.2.8 Biological Shield Model................................................................................ 3-19.

3.2.2.9 Above-Core Components............................................................................. 3-19 3.2.2.9.1 Top Guide............................................................................................. 3-19 3.2.2.9.2 Core*Spray Spargers and Piping.......................................................... 3-19 3.2.2.10 Below-Core Component Models............................................................. :... 3-19 3.2.2.10.1 Lower Fuel Assembly Parts................................................................ 3-19 3.2.2.10.2 Fuel Support Pieces............................................................................ 3-20 3.2.2.10.3 Core Support Plate, Rim Bolts, Dry Tubes, and Bypass Plugs............ 3-20 3.2.2.10.4 Control Blades and Guide Tubes........................................................ 3-20 3.2.2, 11 Summary of the Geometry Modeling Approach.......................................... 3-20 3.2.3 Parametric Sensitivity Analyses........................................................................... 3-21 3.2.4 Particle Transport Calculation Parameters........................................................... 3-22 3.2.5 Fission Spectrum and Neutron Source................................................................. 3-22 3.3 Surveillance Capsule Activation and Fluence Results................................................. 3-22 3.3.1 Summary of the Flux Wire Activation Analysis..................................................... 3-23 3.3.1.1 Summary of the Surveillance Capsule Activation and Fluence Analysis....... 3-23 3.3.2 Comparison of Predicted Activation to Plant.:.specific Measurements................... 3-24 3.3.2.1 Cycle 8 30° Surveillance Capsule Activation Analysis............... :.................. 3-24 3.3.2.2 Cycle 24 120° Surveillance Capsule Activation Analysis.............................. 3-25 3.3.3 Capsule Peak Fluence Calculations and Lead Factor Determinations................. 3-26 3.4 Capsule Fluence Uncertainty Analysis......................................................................... 3-27 3.4.1 Comparison Uncertainty........................................................................................ 3-27 3.4.1.1 Plant-Specific Comparison Uncertainty........................................................ 3-27 3.4.1.2 Benchmark Comparison Uncertainty............................................................ 3-27 X

3.4.2 Analytic Uncertainty............................................................................................. 3-28 3.4.. 3 Combined Uncertainty........................................................................................... 3-28 4 CHARPY TEST DATA....................................................................................... :.................. 4-1 4.1 Charpy Test Procedure................................................................................................. 4-1 4.2 Charpy Test Data for the 120° Capsule.........................................................................4-2 5 CHARPY TEST RESUi.. TS ************************************************************************************************~**5-1 5.1 Analysis of Impact Test Results..................................................................................... 5-1 5.2 Irradiated Versus Unirradiated CVN Properties............................................................. 5-1 6 REFERENCES.............................................*......................................................................* 6-1 A DOSIMETER ANALYSIS............................................................................ :....................... A-1 A.1 Dosimeter Material Description.................................................................................... A-1 A.2 Dosimeter Cleaning and Mass Measurement............................................................... A-1 A.3 Radiometric Analysis.................................................................................................... A-1 XI

(

LIST OF FIGURES Figure 2-1 Drawing Showing the Charpy Test Specimen Geometry and ASTM E23 [9]

Permissible Variations............................. ~.. *........... :... *....................................................... 2-3 Figure 2-2 Photograph of the Hatch Unit 2 120° Capsule (top) and a Magnified View of the External Identification Markings (bottom)....................................,................................ 2-4 Figure 2-3 Photograph of the Hatch Unit 2 120° Capsule..................................................... :... 2-5 Figure 2-4 Photograph of the Inside of the Hatch Unit 2 120° Capsule..................................... 2-6 Figure 2-5 Charpy Energy Plot for Plate Heat C8554 (LT) Unirradiated....................,............ 2..,10 Figure 2-6 Charpy Energy Plot for Weld Heat 51912 Unirradiated......................................... 2-12 Figure 2-7 Lateral Expansion Plot for Plate HeafC8554 (LT) Unirradiated............................. 2-14 Figure 2-8 Lateral Expansion Plot for Weld Heat 51912 Unirradiated..................................... 2-,16 Figure 2-9 Drawing of the Identification Markings Found Inside the Hatch Unit 2 120° Capsule......*... :.....,...................................... *................................................................... 2-19 Figure 2-10 Photograph of the Inside of th.e GS Charpy Packet within,the Hatch Unit 2 120° Capsule..................................................... ;............................................................ 2-20 Figure 2-11 Photograph of the Inside of the G4 Charpy Packet within the Hatch Unit 2 120° Capsule.................................................................................................................. 2-20 Figure 3-1 Planar View of the Hatch Unit 2 Reactor at the Core Mid~Plane Elevation.............. 3-3 Figure 3-2 Planar View of the Hatch Unit 2 Fluence Model at the Core Mid-Plane Elevation in Quadrant Symmetry......... ;....................................................,..................... 3-13 Figure 3-3 Axial View of the Hatch Unit 2 Fluence _Model....................................................... 3-14 Figure 4-1 Illustration of Digital Optical Comparator Measurement of Shear Fracture

.Area............... *.................................... *......................................................................... *...... 4-2 Figure 5-1 Irradiated Plate Heat C8554 Charpy Energy Plot (Hatc,h Unit 2. 120° Capsule) (LT)........................... :....................................................................................... 5-2 Figure 5-2 Irradiated Weld Heat 51912 Charpy Eriergy Plot (Hatch Unit 2 120° Capsule)........ s.,4 Figure 5~3 Irradiated Plate Heat C8554 Lateral Expansion Plot (Hatch Unit.2 120° Capsule) (LT)........ *._....................................... u.............................,........ *........................... 5-6 Figure 5-4 Irradiated Weld Heat 51912 Lateral Expansion Plot (Hatch Unit 2 120° Capsule).........................,...................... :........................................................................... 5-8 Figure A-1 Hat,ch Unit 2 120° Capsule Packet G4 Fe Dosimeter Wire G4 Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right)....................,........................................ A-3 Figure A-2 Hatch Unit 2 120° Capsule Packet G4 Cu Dosimeter Wire G4 Cu: Prior to Cleaning (left); and After Cleaning/Coiling (right)............................................................. A-3 Figure A-3 Hatch Unit 2 120° Capsule Packet G4 Ni Dosimeter Wire G4 Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right),............................................................ A-3 xm I

i

Figure A-4 Hatch Unit 2 120° Capsule Packet GS Fe Dosimeter Wire GS Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right)............................................................. A-4 Figure A-5 Hatch Unit 2 120° Capsule Packet'GS Cu Dosimeter Wire GS Cu: Prior to Cleaning (left); and After Cleaning/Coiling (right)............................................................. A-4 Figure A-6 Hatch Unit 2 120° Capsule Packet GS Ni Dosimeter,Wire GS Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right)............ **************************:**********************A-4 XIV

LIST OF TABLES Table 2-1 Hatch Unit 2 120° Surveillance Capsule Specimen lnventory1......... :................... : *** 2-2 Table 2-2 Best Estimate Chemistry of Available Data Sets for Plate Heat C8554.................... 2-7 Table 2-3 Best Estimate Chemistry of Available Data Sets for Weld Heat 51912..................... 2-7 Table 2-4 Unirradiated Longitudinal CharpyV-Notch Impact Test Results for Surveillance Base Metal (Heat C8554) Specimens from the Hatch Unit 2 Surveillance Program [8].................................................................................................. 2-8 Table 2-5 Unirradiated Charpy V-Notch Impact Test Results for Surveillance Weld Metal (Heat 51912) Specimens from the Hatch Unit2 Surveillance Program [8]..............,......... 2-8 Table 2-6 Baseline CVN Properties........................................................................... '............... 2-9 Table 3-1 Summary of the Hatch Unit 2 Surveillance Capsules and flux Wires....................... 3-1 Table 3-2 Summary of Material Compositions by Region for Hatch Unit 2.............................. 3-5 Table 3-3 Summary of Hatch Unit 2 Core Loading lnventory.................. :................................ 3-8 Table 3-4 State-Point Data for Hatch Unit 2 per Cycle Basis.:................................................. 3-9 Table 3-5 Summary of Fluence and Activity Comparisons for the Hatch Unit 2 Dosimetry..... 3-24 Table 3-6 Comparison of Flux.Wire Calculated-to-Measured Activities for the 30° Surveillance Capsule Removed from Hatch Unit 2 at EOC 8.......... :............................... 3-25 Table 3-7 Comparison of Flux Wire Calculated-to-Measured Activities for the 120° Surveillance Capsule Removed from Hatch Unit 2 at EOC 24........................................ 3-26 Table 3-8 Best-Estimate Fluence and Lead Factors Determined for the Hatch Unit 2 Surveillance Capsules.... :............................................................................................... 3-26 Table 3,..9 Hatch Unit 2 Surveillance Capsule Combined Uncertainty for Energy

>1.0 MeV...... :............,.......,:........................................................................................... 3-29 Table 4-1 Irradiated Charpy V-Notch Impact Test Results for Surveillance Base Metal Specimens (Heat C8554) from the Hatch Unit 2 120° Surveillance Capsule.....................4-3 Table 4-2 Irradiated Charpy V-Notch Impact Test Results for Surveillance Weld Metal

Specimens (Heat 51912) from the Hatch Unit 2 120° Surveillance Capsule.....................4-4 Table 4-3 Irradiated Charpy V-Notch Impact Test Results for Surveillance HAZ Metal Specimens from the Hatch Unit 2120° Surveillance Capsule...........................................4-4 Table 5-1 Effect of Irradiation (E>1.0 MeV) on the Notch Toughness Properties.................... 5-10 Table 5-2 Comparison of Actual Versus Predicted Embrittlement.......................................... 5-11 Table 5-3 Percent Decrease in Upper Shelf Energy............................,.................................. 5-11 Table A-1 Hatch Unit 2 120° Capsule Charpy Packet Dosimeter Wire Masses....................... A-5 Table A-2 Gamma Ray Spectrometer System (GRSS) Specifications.................................... A-5 xv I

I I

Table A-3 Counting Schedule for Hatch Unit 2 120° Capsule Dosimeter Materials................. A-5 Table A-4 Neutron-Induced Reactions of Interest.................................................................... A-6 Table A-5 Results of Hatch Unit 2 120° Capsule Radiometric Analysis................................... A-6 XVI

1 INTRODUCTION Test coupons of reactor vessel ferritic beltline materials are irradiated in reactor surveillance capsules to facilitate evaluation of vessel fracture toughness in vessel integrity evaluations.

The key values that characterize fracture toughness are the reference temperature of nil-ductility transition (RTNDT) and the upper shelf energy (USE). These are defined in 10CFR50, Appendix

G [1] and in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI [2].

Appendix Hof 10CFR50 [1] and ASTM E185-82 [3] establish the methods to be used for testing of surveillance capsule materials.

In the late 1990s, the BWR Vessel and Internals Project (BWRVIP) initiated the BWRVIP Integrated Surveillance Program (ISP) [ 4], and the BWRVIP assumed responsibility for testing and evaluation of ISP capsules. The surveillance plate and weld from the Edwin I. Hatch Nuclear Power Plant (HNPP) Unit 2 (hereinafter, Hatch Unit 2) were designated as "ISP representative surveillance materials" to be tested by the I~P according to an approved capsule withdrawal and test schedule.

This report addresses the withdrawal and testing of the Hatch Unit 2 120° surveillance capsule.

The capsule contained flux wires for neutron flux monitoring, Charpy V-notch impact test specimens, a:nd tensile specimens. The capsule was irradiated for 24 cycles of operation before it was removed in *February 2017 and shipped to MP Machinery & Testing, LLC for opening and testing of the Charpy V-notch surveillance specimens. Evaluation of the fluence environment was conducted by Trans Ware Enterprises, Inc. Final evaluation of the Charpy test data and irradiated.material properties and compilation of this report were performed by EPRI. The Charpy V-notch surveillance materials were tested per ASTM E185-82, and the information and the associated evaluations provided in this report have been performed in accordance with the requirements of 10CFR50, Appendix B [5].

This report compares the irradiated material properties of surveillance plate heat C8554 and surveillance weld heat 51912*to their unirradiated (e.g., baseline) properties. The observed embrittlement ( as characterized by the shift in the Charpy energy curve 3 0 ft-lb ( 41 J) index temperature or ~T30) is compared to that predicted by U.S. Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.99, Revision 2 [6]. Other BWRVIP ISP reports will integrate the results from the 120° surveillance capsule with the results from the Hatch Unit 2 30° surveillance capsule (withdrawn in 1989) for a broader characterization of embrittlement behavior.

1-1

Introduction 1.1 Implementation Requirements The results documented in this report will be utilized by the BWRVIP ISP and by individual utilities to demonstrate compliance with 1 OCFR50, Appendix H, Reactor Vessel Material Surveillance Program Requirements. Therefore, the implementation requirements of 10CFR50, Appendix H govern and the implementation requirements of Nuclear Energy Institute (NED 03-08, Guideline for the Management of Materials Issues [7], are not applicable.

1-2

2 MATERIALS AND TEST SPECIMEN D~SCRIPTION The General Electric (GE) Designed Hatch Unit 2 120° surveillance capsule was removed from the plant and shipped to MP Machinery and Testing, LLC (MPM) for analysis. The capsule contained a total of two Charpy packets and 4 tensile tubes. The 120° degree capsule is an original plant capsule, and has been irradiated in the plant since initial startup. The 30° capsule was tested by GE and the results are reported in Reference [8].

2.1 Dosimeters The dosimetry wires were located along the ends of the Charpy specimens within the Charpy packets during irradiation. Each of the two Charpy packets contained one high i:1urity iron wire, ohe high purity copper wire, and one high purity nickel wire for fluence evaluatibn. Further details on the exact wire locations during the irradiation are provided in the capsule opening discussion given in Section 2.3. A detailed discussion of the radiometric analysis of the capsule dosimetry wires is provided in Appendix A.

2.2 Test Materials The Hatch Unit 2 120° surveillance capsule Charpy V-notch specimen inventory, material descriptions, unirradiated (baseline) Charpy impact data, and previously measured data are summatjzed in this section of the report.

2.2.1 Capsule Loading Inventory The Hatch Unit 2 120° surveillance capsule inventory is provided in Table 2-1. All of the capsule specimens, which include Charpy specimens, tensile specimens, and dosimeters, were recovered from the capsule basket. Testing was performed on all of the 24 Charpy specimens, and the dosimetry wires were counted and weighed to determine specific activities. All eight of the tensile specimens (three base, three weld, and two RAZ) remain untested and are being held in reserve for future surveillance program use. The technical advantage of storing the tensile specimens untested is that there will be options in the future for how these specimens will be used to obtain useful data. For example, the tensile specimen geometry is conducive to fabrication of sub-size Charpy as well as miniaturized Charpy V-notch specimens. Further, research is currently underway to develop testing methods which will enable the determination of plane-strain fracture toughness data from Charpy-sized specimens. With these new technologies in view, there may also be a need in the future for static and/or dynamic tensile data for use in the calculation of fracture toughness from experimental data obtained from Charpy specimens. Therefore, all of the tensile specimens have been placed into the archive storage so that they can be tested when necessary in the future. The broken Charpyspecimen halves have been added to long-term archive storage for future use in mechanical behavior specimen testing, chemistry analysis, and microstructural studies.

2-1

/

Materials and Test Specimen Description As indicated in Table 2-1, there were two Charpy packets in the capsule, and each contained three dosimetry wires (one Fe wire, one Cu wire, and one Ni wire) and 12 Charpy specimens. A drawing of the Charpy test specimen is shown in Figure 2-1 for reference. Photographs of the capsule are given in Figures 2-2 through 2-4. The markings on the outside of the capsule, including the reactor code and the capsule code, were recorded and verified.

Table 2-1 Hatch Unit 2 120° Surveillance Capsule Specimen lnventory1 Charpy Number of Number of Relative Packet Charpy Specimens Flux Wires Vertical No.2 Position Base Weld HAZ Fe Cu Ni Highest G4 8

4 0

1 1

1 Charpy Packet in Basket Lowest GS 0

4 8

1 1

1 Charpy Packet in Basket

1.

The surveillance program includes tensile specimens, but the tensile specimens were not tested.

All eight of the tensile specimens for this capsule were located at axial positions above Charpy packet G4.

2.

The packet numbers in this table are organized by axial position in the capsule with packet G4 at the highest elevation in the reactor and packet GS at the lowest elevation in the reactor.

2-2

Materials and Test Specimen Description I

I

/

~

1_

L

-- 1 0.394 0.079 _l J_~

i: I

! :f o.394 ASTM E23 [9] permissible variations shall be as follows:

Notch length to edge:

Adjacent sides shall be at:

Cross-sectional dimensions:

Length of specimen (L):

Centering of notch (L/2):

Angle of notch:

Radius of notch:

Notch depth:

Finish requirements:

Figure 2-1 90 +/- 2 degrees 90 degrees +/- 10 minutes

+/- 0.003 inches 2.165 (+0.0, -0.100) inches

+/- 0.039 inches

+/- 1 degree 0.010 +/- 0.001 inches

+/- 0.001 inches 63 µ-inch on notched surface and opposite face; 4 µ-inch elsewhere Drawing Showing the Charpy Test Specimen Geometry and ASTM E23 [9] Permissible Variations 2-3

Materials and Test Specimen Description Figure 2-2 Photograph of the Hatch Unit 2 120° Capsule (top) and a Magnified View of the External Identification Markings (bottom)

Figure 2-2 shows the side of the surveillance capsule which faced the reactor vessel. The identification code, "131 C7717G002", was engraved near the hook.

2-4

Materials and Test Specimen Description Figure 2-3 Photograph of the Hatch Unit 2 120° Capsule Figure 2-3 shows the side of the surveillance capsule which faced the reactor core.

2-5

Materials and Test Specimen Description Figure 2-4 Photograph of the Inside of the Hatch Unit 2 120° Capsule 2.2.2 Material Description The Hatch Unit 2 surveillance program is described in a report issued by General Electric in Reference [8]. The Hatch Unit 2 reactor pressure vessel (RPV) is a 218-inch diameter BWR/4 design. The pressure vessel construction was performed by Combustion Engineering to the 1970 Addenda of the 1968 edition of the ASME Code. The pressure vessel shell and head plate materials are ASME A533, Grade B, Class 1 low alloy steel. The nozzles and closure flanges are SA508 Class 2 low alloy steel, and the closure flange bolting materials are ASME SA540 Grade B23 or B24 low alloy steel. The fabrication process employed quench and temper heat treatment immediately after hot forming, then submerged arc welding and post-weld heat treatment. The post-weld heat treatment was typically 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per inch of thickness at 1150 °F. The fabrication impact test specimens were give a simulated post weld heat treatment of 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months .

Base metal specimens were cut from Heat C85 54 and were machined from the 11,i T and % T positions in the plate in the longitudinal orientation (long axis parallel to the rolling direction).

The base metal specimens were stamped with "P 1" on one end designating the base metal and "46" on the other end designating the reactor number. The weld metal and HAZ Charpy specimens were machined from two pieces of the surveillance test plate with a weld which was identical to the beltline longitudinal seam welds. The base metal orientation in the weld and 2-6

Materials and Test Specimen Description HAZ specimems w~s longitudinal. The weld specimens were stamped with "P2:' on one end *

\\

designating the weld metal and 46 on the other end to designate the reactor number. The HAZ specimens were stamped with "P3" on one end destgnating the HAZ metal and 46 on the other

. end to designate the reactor number.

2.2.3 Chemical Composition Table 2-2 details-the best estimate average chemistry values for plate heat C8554 surveillance material. Table 2-3 details the best estimate average chemistry values for weld heat51912 surveillance material. Chemical compositions are presented in weight percent. If there are multiple measurements on a single specimen, those are first averaged to yield a single value for that specimen, and then the different specimens are averaged to determine the heat best estimate.

Table 2-2

. Best Estimate Chemistry of Available Data Sets for Plate.Heat C8554 Cu (wt%)

Ni (wt%)

p (wt%) s (wt%)

Si (wt%)

Specimen ID Source 0.08 0.63 0.011 P1-46-A

. 0.08 0.63 0.009 P1-46-B Reference 8 0.08 0.63 0.010 Average P1-46 0.57 0.010 0.018 0.22 C8554 Slab 1 Reference 1 O 0.. 58 0.010 0.018 0.24 C8554Slab 2 0.08 0.59 0.010 0.018 0.23

+Best Estii:nate Average Table 2-3 Best Estimate Chemistry of Available Data Sets for Weld Heat 51912 Cu Ni p

.S Si Specimen ID Source (wt%)

(wt%)

0.13 0.12 0.013 P2-46~A 0,12 0.07 0.014 P2-46-B Reference 8 0,13 0;10 0.014 Average P2-46 0.13 0.10 0.014

+Best Estimate Average 2.2.4 CVN BaselineProperties Table 2-4 contains the unirradiated Charpy data for the C8554 surveillance plate material. Table 2-5 contains the unirradiated Charpy data for the 51912 surveillance weld material.

2-7

Materials and Test Specimen Description 2-8 Table 2-4 Unirradiated Longitudinal Charpy V-Notch Impact Test Results for Surveillance Base IVletal (HeatC8554) Specimens from the Hatch Unit 2 Surveillance Prowam [8]

Base Unirradiated: Heat C8554, Longitudinal Specimen Test Impact Energy Lateral Percent Temperature Expansion Shear ID OF (OC) ft-lb (J) mils (mm) 1

-100

(-73.3) 5.5 (7.46) 0.5 (0.01) 10 2

-60

(-51.1) 14.0 (18.98) 10.0 (0.25) 17 3

-40

. (-40.0) 26.0 (35.25) 20.0 (0.51) 30 4

-20

(-28.9) 38.5 (52.20) 22.0 (0.56) 32 5

0

(-17.8) 20.0 (27,12) 16.0 (0.41) 18 6

0

(-17.~)

50.0 (67.79) 34.5 (0.88) 35 7

20

(~6.7) 64.0 (86.77) 41.Q (1.04) 49 8

40 (4.4) 82.0 (111.18) 57.0 (1.45) 67 9

60 (15.6) 95.0 (128.80) 63.5 (1.61) 84 10 100 (37.8) 105.0 (142.36) 74.5 (1.89) 100 11 180 (82.2) 114.5 (155.24) 78.0 (1.98) 100 12 400 (204.4) 115.0 (155.92) 76.0 (1.93) 100 Table 2-5 Unirradiated Charpy V-Notch Impact Test Results for Surveillance Weld Metal (Heat 51912)

Specimens from the Hatch Unit 2 Surveillance Program [8]

Weld Unirradi~ted: Heat 51912 Test Impact Energy Lateral Percent Specimen Temper,;1ture Expansion Shear ID OF (OC) ft-lb (J) mils (mm) 1

-100

(-73.3) 6.0 (8.13) 0.0 (0.00) 12 2

-60

(-51.1) 7.5 (10.17) 2.0.

(0.05) 15 3

-40 (AO.O) 12.0 (16.27) 9.5 (0.24) 25 4

-20

(-28.9) 34.0 (46.10) 21.5 (0.55) 44 5

0

(-17.8) 42.0 (56.94) 29.0 (0.74) 28 6

20

(~6.7) 85.5 (115.92) 58.0 (1.47) 70 7

20

(-6.7)

E33.0 (85.42) 40.5 (1.03) 49 8

40 (4.4) 76.0 (103.04) 53.5 (1.36) 72

,9 60 (15.6) 86.0 (116.60) 57.5 (1.46) 76 10 100

. (37.8) 118.5

. (160.66) 85.5 (2.17) 96.

11 180 (82.2) 118.0 (159.99) 88.5 (2.25) 100 12 400 (204.4) 126.0 (170.83) 91.0 (2.31) 100 I

/

Materials and Test Specimen Description The baseline test data were fit to a hyperbolic tangent curve using the computer program CVGRAPH [11]. Figures2-5 and 2-6 show the fitted Charpy energy data curve for the unirradiated plate and weld, respectively. Figures 2-7 and 2-8 show the fitted lateral expansion curve for the unirradiated plate and weld, respectively. Table 2-6 summarizes the unirradiated (baseline) Charpy V-notch properties (index temperatures) of plate heat C8554 and weld heat 51912. In this table and throughout this report, T30 is the 30 ft-lb (41 J) transition temperature; Tso is the ~Oft-lb (68 J) transition temperature; T3smi1 is the 35 mil (0.89 mm) lateral expansion temperature; and USE is the average energy absorption at full shear fracture appearance.

Table 2-6 Baseline CVN Properties Material T30 Tso T3smn Upper Shelf Identity Material OF (OC)

OF (OC)

Energy (USE) ft-lb (J)

C8554 Hatch Unit 2

-19.6 (-28.7) 6.2 (-14.3) 8.1 (-13.3) 111.5 (151.2)

(LT Orientation)

Surveillance Plate 51912 Hatch Unit 2

-21.0 (-29.4) 4.0 (-15.6) 7.3 (-13.7) 120.8 (163.8)

Surveillance Weld 2-9

Materials and Test Specimen Description PLATE HEAT C8554 (HA2)

CV Graph 6.02: Hyperoolic Tangent Cmve Printed on 12/14/2017 4:39 PM A = 57.00 B = 54.50 C = 62.39 TO = 14.25 D = 0.00 Correlation Coefficient= 0.977 Equation is A+ B * [Tanh((T-TO)/(C+D1))]

Upper ShelfEnergy = 111.50 (Fixed)

Lower Shelf Energy= 2.50 (Fixed)

Temp@30 ft-Jbs=-19.60° F Temp@35 ft-lbs=-12.40° F Temp@50 ft-lbs= 6.20°F Plant: Hatch 2 Orientation: LT Material: SA533Bl Capstile: lJNIRRA Heat C8554 Fluence; O.OOE+oOO n/cm2 ti.I

.c -

I

¢:: -

120.,._...............__.............. __....................,...................... __.............. __..................,............................................ __.................

()

0.

1------ -----r---- ------r----- -----T--;r~j

___ -+-

__ ------*r---t-

____ -~-----i~

____ ~_T __ --;-

___ T-t i

f:

i j

[

1------ _____ t" ____ ------:----- ----:-; -----------i------ -----r---- ------r----- -----r**-- ----

80 I:

100

: : J:

r*--- -----r*--- -----r--- --- -T*--:-------r*---- ------1**---- -----r*--- -----r*--- _____ T ___ _

60 -----------t-f-----+--i----+---+--+-----;'-----+--i----+----i-----1 1

I

/_ :

I 1

~------ ------~----- -----.:~----~ -----}-----------.. ~------ -----~------ ------j--.. *-- ------~-----

-200

-100 0

100 200 300 400 500 600 Temperature{° F)

CVGraph 6.02 12/14/2017 Page 1/2 Figure 2-5 Charpy Energy Plot for Plate Heat C8554 (LT) Unirradiated 2-10

Materials and Test Specimen Description Plant: Hatch 2 Orientation: LT Temperature {° F)

-100

-60

-40

-20 0

0 20 40 60 100 180 400.

CVGraph 6.02 Figure 2-5 (Continued)

Material: SA533Bl Capsule: UNIRRA PLATE HEAT C8554 (HA2)

Charpy V-Notch Data InputCVN Computed CVN 5.5 5.2 14.6 11.7 26.0 18.8 38:S 29.8 20.0 44.8 50.0 44.8 64:0 62.0 82:0 78.3 95.0 91.1 105:0 104.9 114.5 11.1.0 115.0 111.5 12/14/2017 Charpy Energy Plot for Plate Heat C8554 (LT) Un irradiated Heat: C8554 Fluence: O.OOE+OOO n/cm2 Differential 0.27 2.27 7.21 8.74

-24.76 5.24 1.99 3.70 3:94 0.06 3.53 3.50 Page2/2 2-11

Materials and Test Specimen Description WELD HEAT 51912 (HA2)

CVGraph 6.02: Hype!bolic Tangent Curve Printed on 12/14/2017 4:43 PM A= 61.65 B = 59.15 C = 62.86 TO= 16.46 D = 0.00 Correlation Coefficient= 0.982 Equation is A+ B * [Tanh((T-TO)/(C+DT))]

Upper Shelf Energy= 120.80 (Fixed)

Lower Shelf Energy= 2.50 (Fixed)

Temp@30 ft-lbs=-21.00° F Ternp@35 ft-lbs=-14.00° F Temp@50 ft-lbs= 4.00°F Plant Hatch 2 Orientation: NA Material: SAW Capsule: UNIRRA Heat 51912 Fluence: O.OOE+oOO n/cm2 140 ---------------------------------.----------------------------------

I I

J I

I I

()

I I

I 120 J--_:*~-1--~*~+--___:_* ---1f--_;__--.,l,.--------:~.:""i-"""""="""'"""""!!!!!!!!!!!l""""'""F""'~""""'"F"""~""""""'I i

'Vu :

i

~ * *.. r-*.. * * *.. * *t--*.. * * * * * *t** *.. *.. *.. t

1*.....

1 * * * * * - * * * * *t *

.. * * * * * * *( :* * * * *

.. *(.. * * *.... t * * *..

i *****~***** *f *of** !

1* * ** **f f *

/

80 t----i-~-1-~.;.----;.~--;.-~+--~~-t-~-i----t~-+-~-t-----i'----+-~-i--~+----+~-1

A

~****1****** ******1****** *:****1***** *i**1***********

60 t-----'-~-t-~.----+~....;.-.~+-1----+~-+-~...;..._--t~-:-~+----i'----+-~-+-~+----+~-I

-*****:******...... ;...... ******:****** '****i************i******...... ;...... ******:****** ******:******......,.....

100 o

O I

,------ ******,-***** ------,----- ------r-**--

I I

I o

I I

1 I

I O

I 40 t---+-:~-t-----,:~-1-~.;...:__,ll'!')~-;:~-t-~+-:~-t-----i:~-t-~+-:~-t-----i:~-+-~+:----1

-1 : : : : : :

- - - 1**** + * +* * *

1

* *+ r ; * :

20 i----;-~-1-~,;----;.~-':-l~+-----;-~-1-~-:----t~-+-~-t-----i---i-+-~-;-~+----,-~-1 JJ.: *****!

L*

f

.f.... J.

01-........ ---+---~---T i

i i

-300

-200

-100 100 200 300 400 500 600 Temperature (0 F)

CVGraph 6.02 12/14/2017 Page 1/2 Figure 2-6 Charpy Energy Plot for Weld Heat 51912 Un irradiated 2-12

Materials and Test Specimen Description Plant: Hatch.2*

Orientation: NA Temperatw:e C° F)

-iOO

~60

-40

-20 0

20 20 40 60 100 180 400 CVGraph 6.02

.Material: SAW Capsule: DN1RRA WELD HEAT 51912 (HA2)

Charpy V-Notch Data InputCVN Computed CVN 6.0 5.3 7.5 12.0 12.0 19.3 34.0 30.T 42:0 46.5 85.5 65.0 63.0 65.0 76.0 82.8 86.0 97.1 l!R5 ID.I 118.0 120.2.

126.0 120.8 12/14/2017 Figure 2-6 (Continued)

Charpy Energy Plot for Weld Heat 51912 Unirradiated j

Heat: 51912 Fluence: O.OOE+OOO n/cm*

Differential 0.66

-4.55

-7.33 3.27

-4.51 20.52 I

CJ.98

-6.82

-11.12 5.45

-2.15 5.20

Page 2/2 2-13

Materials and Test Specimen Description PLATE HEAT C8554 LE (HA2)

CVGraph 6.02: Hype!bolic Tangent Cmve Printed on 6/26/2018 7:32 AM A= 38.60 B = 37.60 C = 62.27 TO= 14.08 D = 0.00 Correlation Coefficient= 0.981 Equation is A+ B * [Tanh((T-TO)/(C+DT))]

Upper ShelfL.E. = 76.20 (Fixed)

Lower ShelfL.E. = 1.00 (Fixed)

Temp@35 mils= 8.10° F Plant: Hatch 2 Orientation: LT Material: SA533Bl Capsule: UNIRRA Heat: C8554 Fluence: O.OOE+oOO n/crn2 v.J -.... e 80.--~~...-~~....... ~~...,....~~--,.~~---.~~~...-~~...... ~~-.-~~

! 0

* ( * * * * *** ***\\** * ** * * * ** * *( * * * * * * * ** t-" ** *c y--T_;_: ---*.-.. +-.-... -. ~--:.. -.

_""'1_ -.. -.

_ *i---.-..

1:c~. ~t-

___.-. ----:--* _-__ -t-_-

__ _ -_ --1 _-_

_-;_

70 i

i /

---**r**-* **-**-:--**-- ******r***** ******1******---*r-**** *****r**** ---**r**** ----**r****- ---*-*:

60 I

i I

_J i

I I

I

:------ -----r---- ******(*** -**1i***------**-r**--* *****:--**** *-**--:--***' ---**r**-- ------i-----

50 t-----+~-+-~+----+~--,-~+--#--l---+~--;-~+----+~-t-~-+----il----;-~-1-~+------1

: : 1: : : : : :

e-

1***** +. f

... +

..... *+.. j +. *+**... +

40 J---.......;....~-1---.;'---1-~...;...._---+~--+-~-l-----.~--+-~.;.....---+~-.--~+---+~-l-~;....---1

l :

I

~----1------ ----*t**--- ------r-----c ----r------------1------ ---**t**--- ---**t**--- ------(*--- ---*--r**---

= 30

~

~ 20 - i *- *r * -TJ * -!-- ----- \\- --

. -- --- L *

f *

-i-- *

-1.

-****(*--- ******!****-- ***-Jr*

0 ***--t*----------*j***--* **-**(**** ****-*!***-*- ------(----- ------t-----

10 t----:--~;----:-~-t--E31,__-+-~-:----+~-:-----1~~~t-----:-~-t-~~-+-~:---1 j

j 7

j j

-*-**r**** *****r~r*-r-**** ******r************r**** *****r-**** *****r**** *****r**** ******:.

0 t:::::i:::::::+/-:=t::...C:b-....i,_-l.._j_.L_....i....--1.___.i.._.J...__j_--1,_,i._..L_j__J

-300

-200

-100 0

100 200 300 400 500 600 Temperature (° F)

CVGraph 6.02 06/26/2018 Page 1/2 Figure 2-7 Lateral Expansion Plot for Plate Heat C8554 (LT) Unirradiated 2-14

Materials and Test Specimen Description Plant: Hatch 2 Orientation: LT Temperature {° F)

-100

-60

-40

-20 0

0 20 40 60 100 180 400 CVGraph 6.02 Figure 2-7 (Continued)

Material: SA533Bl Capsule: UNIRRA PLATE HEAT C8554 LE (HA2)

Charpy V-Notch Data InputL. E.

Computed L. E.

0:5 2.9 IO.ti 7.4 20:0 12:3 22.0 19.9 16.0 30.2 34.5 30.2 41.0 42.2 57.0 53.4 63.5 62,2 74.5 71:7 78.0 75.,8 76.0 76.2 06/26/2018 Lateral Expansion Plot for Plate Heat C8554 (LT) Unirradiated Heat: C85.54 Fluence: O.OOE+OOO n/cm*

Differential

-2.38 2.63 7.74 2.15

-14.24 4.26

-1.16.

3.60 1.30 2.78 2.16

-0.20 Page 2/2 2-15

/

Materials and Test Specimen Description WELD HEAT 51912 LE (HA2)

CVGraph 6.02: Hyperoolic Tangent Cmve Printed on 6/26/2018 7:47 AM A = 44.65 B = 43.65 C = 65.52 TO = 22.02 D = 0.00 Correlation Coefficient = 0.984 Equation is A+ B * [Tanh((T-TO)/(C+DT))]

Upper ShelfL.E. = 88.30 (Fixed)

Lower Shelf L.E. = 1.00 (Fi-.,;ed)

Temp@35 mils= 7.30° F Plant Hatch 2 Orientation: NA Material: SAW Capsule: UNIRRA Heat 51912 Fluence: O.OOE+oOO n/cm' 100 I

I I

I O

I I

O 1'

t l

I O

I I

I 90 80

~

70

-*-!3

= 60 0 *-

~ =

= 50

~

~ -

40

=

~

n

_u.,n,i.-~*--r-~--;--~*--i--~---t

-*--**:**---* **-**{-**- ------[----; --*---(----C7'f-*-- ****)***-- ***-*t---** -****)****- **-***!*-*--*

/

--*--(-*- *****-:----** -----*(*** ----*-;--* ------*-1**---* ---*-(**-* ------:----*- ---*-*(*** *--*-*r**---

-*-*-r--- ------1------ -----+---- ------1 ---------+----- ----+----- ------~----- -----+---- ------f -----

-*-**(-*** *****{**** -**--+-*-- -~~~---*--*--*i--***- **-*+***-* **--*-!--**** ****-+**** -*****f ****-

---*--l**---- ------!------ --**--!---** *i-+-**--*---*+*-*** *-*-+***** ***---!**---- --****f ***-* *-**-+-*--

: : r:

1------ -----+----- ------r----- -----t------------~------ -----~------ -----+-.. --- ------r----- ------:-----

30

=

~

I l

I f

I I

O I

l I

f I

I I

I 20 -T----(/' :-

1----- --L---- -----------

-t--- -----:--

(---* ****-+**-*- ****--rz--* ----*+-*--****-*+**-** **-*+**--- --*---:---**- -**---~----- *****-r---*-

10 0 -+ -+~~----+ -+-- +- +--+-

~ ~ ~

0

~

Temperature (0 F)

CVGraph 6.02 06/26/2018 Page 1/2 Figure 2-8 Lateral Expansion Plot for Weld Heat 51912 Unirradiated 2-16

Materials and Test Specimen Description Plant: !latch 2 Orientation: NA Temperature {° F)

-TOO

-60

-40

-20 0

20.

20 40 60 100 180 400 CVGraph 6.02 Figure 2-8 (Continued)

Material: SAW Capsule: lJNIRRA WELD HEAT 51912 LE (HA2)

Charpy V-Notch Data InputL. E.

Computed LE.

O:O 3.1 2:0 7.6 9.5 12.4 21.5 20.0 29.0 30.5 58.0.

43.3 40.5 43.3 53.5 56.3 57.5 67.5 85.5 80.9 88.5 81.6 91.0 88.3 06/26/2018

Lateral Expansion Plot for Weld Heat 51912 Unirradiated Heat; 519U Fluence: O.OOE+OOO n/cm2 Differential

-3.06

-5.60

-2.92 1.55

-1.51 14.70

-2.80

-2.83

-9.95 4.59 0.90 2.70 Page 2/2 2-17

Materials and Test Specimen Description 2.3 Capsule.Opening As shown in Figures 2-2 through 2-4, the 120° capsule consisted of a container holding two Charpy packets and four tensile tubes. Each Charpy packet contained 12 Charpy specimens and each tensile tube contained 2 tensile specimens. The outside of the capsule had identification markings which could be clearly read. On one side, the capsule container was marked with the reactor and capsule codes. The reactor code 46 matches the reactor code in Reference [8]. The capsule container was engraved with the marking "131 C7717G002" on the side facing away from the core.

Attention was paid to the location of the Charpy packets, specimens and dosimetry wires during disassembly of the capsule. The dosimetry wire location along the ends of the Charpy specimens are shown in Figure 2-9. Referring to the figure, the 8 base metal specimens and 4 weld specimens in the G4 Charpypacket were installed near the middle of the capsule and the 8 HAZ specimens and 4 weld specimens in the G5 Charpy packet were installed in the bottom of the capsule. Specimen orientation can be seen in Figure 2~9 and the inside of the packets is shown in Figures 2-10 and 2-11. The specimens were given number identifiers 1 to 24 with a"-" in front of the number to uniquely identify each specimen. This can be seen in Figure 2-9. The dosimetry wires and Charpy specimens were placed in individually marked containers for positive identification throughout the work.

2-18

GI -

46 -

131C7716 a G4 a

REACTOR 46 131C7716 a G5 REACTOR 46 a

r--

i--

1--

i--

Figure 2-9 I P l

-1 Base 11 Weld I P3

-5 HAZ I

P2

-7 Weld 46 Pl Pl Pl

[Base Base Base Base 0

-13

-14

-15

-16 Pl 46 46 46 P3 P3 P3 46 0

HAZ HAZ IHAZ iHAZ

-I

-2

-3

-4 46 46 46 P3 REACTOR NO.

46 I I Pl P2 I I 46 46 I 1 46 46 I I Pl Pl Pl Base Base

-17

-18 46 46 46 46 HAZ HAZ

-5

-6 P3 P3 Materials and Test Specimen Description 46 I Base

-2 Weld P2 I

-02

-4 HAZ P3 I-

-6 G3 Base 46 I- -

G4

-8 46 46 46 P2 P2 P2 V-notch facing down Base [Base Weld Weld Weld Weld /

/

3 dosimetry wires

-19

-20

-21

-22

-23

-24 in Charpy packet

/

Pl Pl P2 46 46 46

/

P3 P3 P2 P2 P2 P2 V-notch facing down HAZ HAZ Weld Weld Weld Weld V /

3 dosimetry wires

-7

-8

-9

-10

-II

-12 in Charpy packet

/

46 46 46 46 46 46

/

131C7717G002 Note:

Specifies that markings are a

stamped on the back side of the packet that the arrow is pointing to.

IDs Pl, P2, P3 or 46 were stamped on the ends of the specimens.

Drawing of the Identification Markings Found Inside the Hatch Unit 2 120° Capsule 2-1 9

Materials and Test Specimen Description Figure 2-10 Photograph of the Inside of the GS Charpy Packet within the Hatch Unit 2 120° Capsule Figure 2-11 Photograph of the Inside of the G4 Charpy Packet within the Hatch Unit 2 120° Capsule 2-20

3 NEUTRON FLUENCE CALCULATION This section presents the results of an analysis to determine the dosimetry activity and fast neutron fluence in surveillance specimens and flux wires removed from the Hatch Unit 2 reactor.

The computed activities are compared to measurements for all pressure vessel surveillance capsules and flux wires irradiated in the Hatch Unit 2 reactor up to and including the withdrawal of the 120° surveillance capsule from the reactor at the End of Cycle (EOC) 24. This surveillance capsule, which included copper, iron and nickel flux wire dosimetry specimens, had achieved a lifetime exposure of 30.3 effective full power years (EFPY).

In addition to the end of Cycle 24 (EOC 24) capsule evaluation, activities, and fluence were determined for a surveillance capsule removed at the end of Cycle 8 (EOC 8) from the Hatch Unit 2 reactor. Table 3-1 summarizes the surveillance capsules evaluated in this report for the Hatch Unit 2 reactor.

Table 3-1 Summary of the Hatch Unit 2 Surveillance Capsules and Flux Wires Capsule Location Time of Removal EFPY at Removal Reference 30° EOC 8 6.61

[8]

120° EOC 24 30.3 The determination of activation, fluence and combined uncertainty for the surveillance capsules and flux wires listed in Table 3-1 is based upon the RAMA Fluence Methodology [12], the RAMA Fluence Methodology Procedures Manual [13], and the RAMA Fluence Methodology Theory Manual [ 14].

The RAMA Fluence Methodology (hereinafter referred to as "RAMA") was developed by Trans Ware Enterprises Inc. under sponsorship of EPRI and the BWRVIP. In accordance with the requirements of U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [15],

RAMA is qualified against industry standard benchmarks for both boiling water reactor (BWR) and pressurized water reactor (PWR) designs. The RAMA methodology and Trans Ware's application of the methodology have been reviewed by the NRC and given generic approval for determining fast neutron fluence in both BWR and PWR pressure vessels [ 16] with no discemable bias in the computed results.

The Hatch Unit 2 reactor is a BWR/4 class design with a core configuration of 560 fuel assemblies. The RAMA methodology has been further qualified against 192 plant-specific dosimetry measurements for BWR/4 class plants of multiple core configurations and is shown to have an overall unbiased computed-to-measurement (C/M) ratio and standard deviation of 1.02 +/- 0.11.

3-1

Neutron Fluence Calculation 3.1 Description of the Reactor System This section provides an overview of the reactor design and operating data inputs that were used to develop the computational fluence model for the Hatch Unit 2 reactor. All reactor design and operating data inputs used to develop the model are plant-specific and were provided by Southern Nuclear Operating Company, Inc. (SNOC). The inputs for the fluence geometry model were developed from nominal and as-built drawings for the RPV, RPV internals, fuel assemblies, and containment regions. The reactor operating data inputs were developed from core simulator data and other sources, as available, that provide a historical accounting of how the reactor operated for Cycles 1 through 24.

3. 1. 1 Overview of the Reactor System Design Hatch Unit 2 is a General Electric BWR/4 class reactor with a core loading of 560 assemblies.

The unit began commercial operation in 1979 with a design rated power of 2436 MWt. Power uprates were achieved in Cycle 13 raising the power to 2558 MWt, in Cycle 15 raising the thermal power output to 2763 MWt, and in Cycle 18 raising the thermal power output to 2804 MWt. At the time of this fluence evaluation, Hatch Unit 2 has completed 24 cycles of operation.

Figure 3-1 illustrates the basic planar configuration of the Hatch Unit 2 reactor at an axial elevation near the reactor core mid-plane. All the radial regions of the reactor that are required for fluence evaluations are shown. Beginning at the center of the reactor and projecting outward, the regions include: the core region; core reflector region (bypass water); central shroud wall; downcomer water region including the jet pumps; RPV wall; cavity region between the RPV wall and insulation; insulation; cavity region between the insulation and biological shield; and the biological shield wall. Cladding is included on the inner RPV surface as well as the inner and outer surfaces of the biological shield wall. Also represented in Figure 3-1, are notations indicating the control rod and fuel assembly locations within the core. Note that the fuel locations are shown only for the northeast quadrant of the core region.

3. 1.2 Reactor System Mechanical Design Inputs The mechanical design inputs used to construct the Hatch Unit 2 fluence geometry model are based upon nominal design and as-built dimensional information. As-built data is always preferred when constructing plant-specific reactor fluence models; however, as-built data is not always available and nominal dimensions are used.

For the Hatch Unit 2 fluence model, the predominant dimensional information used to construct the fluence model is nominal design data. As-built dimensional data was used for the following reactor components:

Instrumentation nozzle (N16) centerline azimuth Shroud lower wall outer radius (at lower elevation)

Shroud lower flange inner radius Shroud central wall inner radius Shroud head flange inner radius 3-2

Neutron Fluence Calculation An important component of a computational RPV fluence model is the accurate description of the surveillance capsules installed in the RPV. Figure 3-1 shows that the Hatch Unit 2 reactor was initially equipped with three surveillance capsules. The capsules were installed at an elevation around the reactor core mid-plane. Each capsule was mounted radially near the inside surface (OT) of the RPV wall. The surveillance capsules were distributed around the circumference of the RPV at the 30°, 120° and 300° azimuths relative to the reactor north 0° angular direction. After withdrawal of the 30° capsule from the RPV, a reconstituted capsule was inserted into the open location. All capsules removed from the Hatch Unit 2 reactor are evaluated in this report.

Reactor North 315° 300° 240° 225° 210° Notes: This drawing is not to scale.

F = Fuel bundle locations. (Locations shown only for the northeast quadrant.)

+ = \\.nntrnl rM Jocr1tinn i::.

Figure 3-1 a*

1ao*

30° Surveillance Capsule Shroud Repair Tie Rod so*

120° Core Reflector Downcomer Reactor Pressure Vessel and Cladding (Inside)

Thermal Insulation Jet Pump Assembly Biological Shield (Concrete Wall) and Cladding Planar View of the Hatch Unit 2 Reactor at the Core Mid-Plane Elevation 3-3

Neutron Fluence Calculation

3. 1.3 Reactor System Material Compositions Each region of the reactor is comprised of materials that include reactor fuel, metal, water, insulation, concrete, and air. Accurate material information is essential for the fluence evaluation as the material compositions determine the scattering and absorption of neutrons throughout the reactor system and, thus, affect the determination of neutron fluence in the RPV, surveillance capsules, RPV internals, and ex-vessel structures.

Table 3-2 provides a summary of the materials for the principle components and regions of the Hatch Unit 2 reactor. The material attributes for the metal, insulation, concrete, and air compositions (i.e., material densities and isotopic concentrations) are assumed to remain constant for the operating life of the reactor. The bulk water coolant properties throughout the reactor system, with the exception of the core region, are determined assuming rated power and flow conditions. The coolant properties remain constant unless there is a reported change in system heat balance conditions that affect the water properties in the reactor. The nuclear fuel compositions and coolant properties in the reactor core region change continuously during reactor operation. The fuel and coolant properties in the core region are updated for each reactor statepoint condition based on the actual or predicted operating states of the reactor. Water properties immediately above and below the core region are updated on a cycle-by-cycle basis based on average cycle operating conditions.

3-4

Neutron Fluence Calculation Table 3-2 Summary of Material Compositions by Region for Hatch Unit 2 Region Mat~rial Composition Biological Shield Clad Carbon Steel Biological Shield Wall.

Reinforced Concrete Cavity Regions Air Control Rod Guide Tubes Stainless Steel Control Rods Stainless Steel, B4C Core Outlet.

Steam

Core Reflector Water Core Spray Sparger Flow Areas Water Core Spray Sparger Pipes Stainless Steel Core Support Plate Stainless Steel Core Support Plate Rim Bolts Stainless Steel Downcomer Region Water Fuel Assembly Lower Tie Plate Stainless Steel, Zircaloy, lnconel, Water Fuel Assembly Upper Tie Plate Stainless Steel, Zircaloy, Water Fuel Support Piece Stainless Steel Insulation Stainless Steel, Aluminum, Air Insulation Clad Stainless Steel.

Jet Pump Hold Down Beams lnconel Jet Pump Hold Down Brackets Stainless* Steel Jet Pump Riser and Mixe~ Flow Areas Water Jet Pump Riser and Mixer M~tal Stainless Steel Jet Pump Riser Brace Stainless Steel

. Jet Pl.imp Riser Brace Pads lncoriel

Reactor Coolant / Moderator Water Reactor Core 235LJ, 23BLJ, 239Pu, 240Pu, 241Pu, 242Pu, 0fuel, Zircaloy Reactor Pressure Vessel Clad Stainless Steel Reactor Pressure Vessel Nozzle Forgings Carbon Steel Reactor Pressure Vessel Wall Carbon Steel Shroud Stainless Steel Shroud Repair Tie Rod

.. Stainless Steel Steam Generator Stand Pipes Stainless Steel Surveillance Capsule Specimens Carbon Steel Tie Rod Hardware lnconel

Top Guide.

Stainless* Steel 3-5

Neutron Fluence Calculation

3. 1.4 Reactor Operating Data Inputs An accurate evaluation ofRPV and component fluence requires an accurate accounting of the reactor's operating history. The principle operating parameters that affect the determination of neutron fluence in light water reactors include: core configurations and fuel assembly designs, power history, exposure and isotopic distributions, and water density distributions. The following subsections provide additional information on the characterization of reactor operating data for fluence evaluations.

3.1.4.1 Core Configuration and Fuel Designs The reactor core configuration and the fuel assembly designs loaded in the reactor determine the neutron source and spatial source distribution contributing to the irradiation of the RPV, RPV internals and ex-vessel supporting structures. The Hatch Unit 2 core consists of 560 assemblies.

Figure 3-1 shows the loading configuration of the fuel assemblies in the reactor.

It is common in BWRs that more than one fuel assembly design may be loaded in the reactor core in any given operating cycle. fu order to determine accurate spatial fluence profiles throughout the reactor system, it is important to account for the different fuel assembly designs loaded in the reactor over the operating lifetime of the reactor, especially those designs that reside in the peripheral locations of the core region.

Table 3-3 provides a summary of the many fuel assembly designs that have been loaded in the Hatch Unit 2 reactor core for each operating cycle evaluated in this report. Table 3-3 also identifies the dominant fuel design loaded on the core periphery for each cycle.

3.1.4.2 Reactor Power History Reactor power history is the measure of reactor power levels and core exposure on a continual or periodic basis. For this fluerice evaluation, the power history for the Hatch Unit 2 reactor was developed from daily power history inputs provided by SNOC for operating Cycles 1 through

24. The power history for the Hatch Unit 2 reactor also accounts for periods of reactor shutdown due to refueling outages and ~ther events that affect the activation and decay of dosimetry data.

Table 3-4 provides a summary of the operating history of the Hatch Unit 2 reactor for each operating cycle. The number of statepoints used to represent the operating history and core power distributions of the reactor are listed for each cycle.

Table 3-4 also shows that the reactor operated with a design rated thermal power of2436 MWt for Cycles 1 through 12. The reactor implemented a power uprate in Cycle 13 bringing the rated thermal power to 2558 MWt at which it operated until Cycle 14. The power was then uprated to 2763 MWt until Cycle 18 when it was uprated to 2804 MWt where it now remains.

Table 3-4 further shows the accumulated EFPY at the end of each cycle. The accumulated EFPY computed from the operating data provided by SNOC was verified against power production and exposure records obtained*separately for the plant.

3.1.4.3 Reactor State-Point Data Statepoints break up operating history into ranges of operation based on similar power, exposure, and isotopic distributions. Typically, several statepoints are chosen for each cycle to represent 3-6

Neutron Fluence Calculation different operating conditions experienced by the reactor over the course of that cycle. Table 3-4

shows that a total of 234 statepoints are used to represent the operating history of the Hatch Unit*

2 reactor for the first 24 cycles of reactor operation. It is also shown that the number of statepoints varies appreciably between the cycles. This variation is due to ~everal factors but is mostly related to the availability of data to represent the operating conditions of the reactor for any given operating cycle.

Table 3-4 shows that very' few statepoints were used for each of Cycles 1 through 9. Detailed electronic data was not available for these cycles. The operational histories for these cycles were approximated by Trans Ware using information derived from.best-available paper sources including Supplemental Reload Licensing Submittals (SRLS), Cycle Management Reports (CMR), and Cycle Summary Reports (CSR) for Hatch Unit 2. These reports collectively_

contained core loading patterns, cycle exposure maps, core-average axial power shapes, and control rod patterns for the cycles.

Core simulator data was provided in electronic form by SNOC to characterize the historical operating conditions of the Hatch Unit 2 reactor for Cycles 10 through 24. Core simulator. data represents the best-available information about the reactor's operating history when available.

The data provided for Cycles 10 through 14 was derived from older core simulator methods that included only nodal power histories. The data provided for Cycles 15 and later was derived from modern*core simulator methods that included pin power reconstruction techniques. The core simulator data that was provided by SNOC was processed by Trans Ware to produce statepoint data files for input to the fluence model.

A separate neutronics transport calculation is performed for each selected statepoint. The neutron fluxes calculated for eachl statepoint are then combined with the appropriate daily power history data described in Section 3.1.4.2 to provide an accurate accounting of the neutron fluence for the RPV, RPV internals, and surveillance capsules. The periods of reactor shutdown are also accounted for in this process, particularly to allow for an accurate calculation of irradiated surveillance capsule activities.

3-7

Neutron Fluence Calculation Table 3-3 Summary of Hatch Unit 2 Core Loading Inventory 8x8 Fuel Assembly 9x9 Fuel Assembly 10x10 Fuel Dominant Cycle Designs Designs Assembly Peripheral Designs Fuel GE6/7 GE8 GE9 GE11 GE13 GE14 GNF2 Design 1

560 GE6/7 2

560 GE6/7 3

560 GE6/7 4

560 GE6/7 5

560 GE6/7 6

560 GE6/7 7

560 GE6/7 8

560 GE6/7 9

560 GE6/7 10 384 172 4

GE6/7 11 212 344 4

GE6/7 12 56 58 442 4

GE9 13 556 4

GE9 14 376 184 GE9 15 184 372 4

GE9 16 556 4

GE13 17 556 4

GE13 18 328 232 GE13 19 88 472 GE13 20 560 GE14 21 560 GE14 22 556 4

GE14 23 556 4

GE14 24 332 228 GE14 3-8

Neutron Fluence Calculation Table 3-4 State-Point Data for Hatch Unit 2 per Cycle Basis Cycle Number Number of Reactor Rated Thermal Power 1 Accumulated EFPY Statepoints (MWt) 1 1

2436 1.05 2

1 2436 1.84 3

1 2436 2.49 4

1 2436 2.89 5

1 2436 3.39 6

1 2436 4.47 7

4 2436 5.39 8

7 2436 6.61 9

2 2436 7.78 10 11 2436 9.00 11 9

2436 9.97 12 13 2436 11.3 13 12 2558 12.5 14 11 2558 13.9 15 16 2763 15.1 16 15 2763 16.5 17 13 2763 17.8 18 16 2804 19.5 19 21 2804 21.4 20 15 2804 23.2 21 16 2804 25.0 22 I

13 2804 26.6 23 17 2804 28.5 24 17 2804 30.3

1. The rated thermal power for the reactor is listed for each cycle. Note, however, that the actual daily power levels are used in the determinations of fluence and surveillance capsule activities.

3.1.4.4 Reactor Coolant Properties The reactor coolant water densities used in the fluence model are determined using combinations of core simulator codes and reactor heat balance data.

The water densities used in the reactor core region are derived directly from the thermal-hydraulic calculations performed by core simulator codes. In general, core simulator codes provide active and bypass flow water density data for each fuel assembly in the core. Coolant densities are provided incrementally along the axial height of the assembly as the coolant flows from core inlet to core exit. The water density in the reflector region between the core and core shroud are determined using the average pressure of the core region assuming the water is unvoided saturated liquid.

3-9

I I.

I Neutron Fluence Calculation The water densities above the core, and specifically in the shroud upper plenum region, assume the steam quality exiting the core. There is no mixing of the exit steam with the unvoided or slightly voided bypass flow also exiting the core. This treatment of core exit water densities provides conservative conditions for determining fluence in the upper vessel components and for determining the elevation for the extended RPV beltline. If steam separator stand pipes are present in the model, the core exit steam is also used to fill the steam separator stand pipes that extend above the shroud head.

The core spray sparger piping, which is present in the upper shroud plenum region, is filled with saturated water on the inside of the piping and is surrounded by the plenum core exit steam on the outside of the pipes.

The bulk water densities in the other regions of the RPV are determined from plant-specific heat balance data. The water densities that are calculated in this manner include the core inlet, downcomer, jet pump flow, and feedwater. Heat balance data provides water properties in terms of temperature, pressure, and enthalpy assuming 100% power and flow operating conditions. The water densities determined at full power conditions remain constant throughout the cycle, and for all power states of the reactor throughout a cycle, which should be bounding for best-estimate fluence predictions.

3.2 Methodology This section provides an overview of the methodology and modeling approach used to determine fast neutron fluence for the Hatch Unit 2 RPV and RPV internals, and the fast neutron fluence and activations for the reactor surveillance capsules. The fluence model for Hatch Unit 2 is a plant-specific model that is constructed from the design inputs described in Section 3.1. The computational tools used in the fluence and activation analyses are based on the RAMA Fluence Methodology (RAMA) software [12]. The RAMA Fluence Methodology is described in the RAMA Theory Manual [14]. A general approach for using the toolset is presented in the RAMA Procedures Manual [ 13].

3.2. 1 Computational Method The RAMA Fluence Methodology is a system of computer codes, a.data library, and an uncertainty methodology that determines best-estimate fluence and activations in light water RPVs and RPV internals. The primary software that comprises the methodology includes modd builder codes, a particle transport code, and a fluence calculator code.

The primary inputs for the fluence methodology are mechanical design parameters and reactor operating history data. The mechanical design inputs are obtained from plant-specific design drawings, which include as-built measurements when available. The reactor operating history data is obtained from multiple sources, such as core simulator software, system heat balance calculations, daily operating logs, and cycle summary reports. A variety of outputs are available from the fluence methodology that include neutron flux, fast neutron fluence; dosimetry activation, and an uncertainty analysis.

The model builder codes consist of geometry and material processor codes that generate input for.

the RAMA transport code. The geometry model builder code uses mechanical design inputs and meshing specifications to generate three-dimensional geometry models of the reactor. The 3-10

Neutron Fluence Calculation material processor code uses reactor operating data and material property inputs to process fuel materials, structural materials, and water densities that are consistent with the geometry meshing generated by the geometry model builder code.

The RAMA transport code performs three-dimensional neutron flux calculations using a deterministic, multigroup, particle transport theory method with anisotropic scattering that is based upon the Method of Characteristics [17]. The transport solver is coupled with a general geometry modeling capability based on combinatorial geometry techniques. The coupling of general (arbitrary) geometry with a deterministic transport solver provides a flexible, efficient, and stable method for calculating neutron flux in light water RPVs, RPV components, and structures. The primary inputs for the transport code include the geometry and material data generated by the model builder codes and numerical integration and convergence parameters for the iterative transport calculation. The primary output from the transport code is the neutron flux in multigroup form for every material region mesh in the fluence model.

The fluence calculator code determines fluence and activation in the RPV, surveillance specimens, and RPV components over specified periods of reactor operation. The fluence calculator also includes treatments for isotopic production and dec~y that are required to calculate specific activities for irradiated materials, such as the dosimetry specimens in the surveillance capsules. The primary inputs to the fluence calculator include the multigroup neutron flux from the transport code, response functions for the various materials in the reactor, reactor power levels for the operating periods of interest, specification of which components to evaluate, and the energy ranges ofinterest for evaluating neutron fluence. The reactor operating history is generally represented with several reactor statepoints that represent the core power and core power distributions of the reactor over the operating life of the reactor. These statepoints are integrated with the daily variations in reactor power levels to predict the fluence and activations accumulated throughout the reactor system.

The RAMA nuclear data library contains atomic mass data, nuclear cross-section data, response functions, and other nuclear constants that are needed for each of the code tools. The structure and contents of the data contained within the nuclear data file are based on the BUGLE-96 nuclear data library [18], with extended data representations derived from the VITAMIN-B6 data library [19].

The uncertainty methodology provides an assessment of the overall accuracy of the fluence and activation calculations. Variations in the dimensional data, reactor operating data, dosimetry measurement data, and nuclear data are evaluated to determine if there is a statistically significant bias in the calculated results that might affect the determination of the best-estimate fluence for the reactor. The plant-specific results are also weighted with comparative results from experimental bern;hmarks and other plant analyses and analytical uncertainties pertaining to the methodology to determine if the plant-specific model under evaluation is statistically acceptable as defined in Regulatory Guide 1.190 [ 15].

3.2.2 Fluence Model Section 3.1 describes the design inputs that were provided by SNOC for the Hatch Unit 2 capsule fluence and activation evaluation. These design inputs are used to develop a plant-specific, three-dimensional computer model of the Hatch Unit 2 reactor.

3-11

Neutron Fluence Calculation Figures 3-2 and 3-3 provide general illustrations of the primary components, structures, and regions developed for the Hatch Unit 2 fluence model. Figure 3-2 shows the planar configuration of the reactor model at an elevation corresponding to the reactor core mid-plane elevation. Figure 3-3 shows an axial configuration of the reactor model. Note that the figures are not drawn to scale and do not include representations of the meshing developed for this evaluation. The figures are intended only to provide a perspective for the layout of the model, and specifically how the various components, structures, and regions lie relative to the.reactor core region (i.e.,

the neutron source). Additional detail is beyond the scope of this document.

3-12

Biological Shield (Concrete W all) and Cladding 30° Neutron Fluence Calculation Surveillance Capsule 26 25 24 23 22 21 20 19 18 17 16 15 ii= 14 l'---t'---"l~"'I'---*<--,,

l-+--+--11-+--¥~1<----x---:,,c--,

.~-:...

ii=

14 15 16 17 18 19 20 21 22 23 D Interior Fuel Assemblies

~ Peripheral Fuel Assemblies Noles: This drawing is not to scale, Dimensions are giv en in inches (cm),

Figure 3-2 Shroud Core Reflector Reactor Pressure V essel and Cladding (Inside) 45° Thermal Insulation 6QO Cavity Downcomer Planar View of the Hatch Unit 2 Fluence Model at the Core Mid-Plane Elevation in Quadrant Symmetry 3-13

Neutron Fluence Calculation Tcpof... AuoncoModel Shroud Upper Plenum (Steam Dome) eo.,

Boaomol... ~-

Figure 3-3 l

a:

J

d Axial View of the Hatch Unit 2 Fluence Model 3.2.2.1 Geometry Model j

Jr -~

!i 1 J m

I a:

£ j The Hatch Unit 2 fluence model is constructed on a Cartesian coordinate system using a generalized three-dimensional geometry modeling technique based on combinatorial geometry.

An axial plane of the reactor model is defined by the (x,y) coordinates of the modeling system.

The axial elevation at which a plane exists is defined along a perpendicular z-axis of the 3-14

Neutron Fluence Calculation modeling system. This allows any point in the reactor model to be referenced by specifying the (x,y,z) coordinates, for that point.

The geometry modeling system allows for solid body elements to be constructed to describe the various components, regions, and structures of the reactor. The primary elements used to construct the fluence model include rectangular parallelepip~ds, right circular cylinders, rotatable boxes, truncated cones, spheres, approximated toroids, and wedges. This modeling approach permits a model to be developed in any level of high-definition detail, such as is necessary for fluence and activation evaluations.

Figure 3-1 illustrates a planar cross-section view of the Hatch Unit 2 reactor design at an axial elevation corresponding to the reactor core mid-plane. It is shown for this one elevation that the reactor design is a complex geometry composed of various combinations ofrectangular, cylinc,lrical, and wedge-shaped bodies. When the reactor is viewed in three dimensions, the varying heights of the different components, structures, and regions create additional geometry modeling complexities. An accurate representation of these geometrical complexities in a predictive computer model is essential for calculating accurate, best-estimate fluence in the RPV, surveillance capsules, RPV internals, and the supporting structures inside and outside of the RPV.

Figure 3-2 and Figure 3-3 provide general illustrations of the planar and axial geometry complexities that are represented in the fluence model. For comparison purposes, the planar view illustrated in Figure 3-2 corresponds to the core elevation illustrated in Figure 3-1. The fluence model assumes reflective azimuthal quadrant symmetry in the planar dimension.

Figure 3-2 illustrates the quadrant geometry that is modeled in this analysis. In terms of the modeling coordinate system, this quadrant is designated the "northeast" quadrant of the reactor system. The 0-degree azimuth, which has a "north" designation, corresponds to the 0-degree azimuth referenced in the plant drawings for the RPV. Azimuthal position is incremented clockwise, resulting in the 90-degree azimuth being designated as the "east" direction.

Figure 3-3 illustrates the axial configuration of the primary components, structures, and regions in the fluence model. The figure shows that the axial height of the fluence model spans from a lower elevation below the recirculation nozzles to above the core shroud head. Although not explicitly shown in the Figure 3-3, the model includes the recirculation outlet nozzles.

As previously noted, Figure 3-2 and Figure 3-3 are not drawn precisely to scale and are intended only to provide a perspective of how the various components, structures, and regions of the reactor are positioned relative to the reactor core region. The following subsections provide additional information on the constituent models developed for the individual components, structures, and regions of the fluence model.

3.2.2.2 Reactor Core and Core Reflector The reactor core contains the nuclear fuel that is the source of the neutrons that irradiate all components and structures of the reactor. The core is surrounded by a shroud structure that serves to channel the reactor coolant through the core region during reactor operation. The coolant-containing region between the core and the core shroud is the core reflector. The reactor core geometry is rectangular in design and is modele.d with rectangular elements to preserve its shape in the analysis. The core reflector region interfaces with the rectangular shape of the core 3-15

Neutron Fluence Calculation region and the curved shape of the core shroud. It is, therefore, modeled using a combination of rectangular and cylindrical elements.

The core region is centered in the RPV and is radially-characterized in the analysis with two fundamental fuel zones: interior fuel assemblies and peripheral fuel assemblies. The peripheral fuel assemblies are the primary contributors to the neutron source in the fluence calculation.

These assemblies are loaded at the core edge where neutron leakage from the core is greatest. As a result, there is a sharp power gradient across these assemblies that requires consideration. To account for the power gradient, the peripheral fuel assemblies are sub-meshed with additional elements that approximate the pin-wise details of the fuel assembly geometry and power distribution within the pins. ;The interior fuel assemblies make a lesser contribution to the reactor fluehce and are, therefore, modeled in various homogenized forms in accordance with their contributions to the reactor fluence. For computational efficiency, homogenization treatments are used in the interior core region primarily to reduce the number of mesh regions that must be solved in the transport calculation. The meshing configuration for each fuel assembly location in the core region is determined by parametric studies to ensure an accurate estimate of neutron flux and fluence throughout all regions of the reactor system.

Each fuel assembly design is axially-characterized with five axial material zones: the lower fuel nose piece zone, the lower tie plate/end plug zone, the fuel zone, the fuel upper plenum zone, and the upper tie plate/end plug zone. The structural materials in the top and bottom regions for each unique assembly design are represented in the model to address the shielding effects that these materials have on the components above and below the core region. The fuel zone contains the nuclear fuel and structural materials for the fuel assemblies. Each fuel location in the reactor core is represented uniquely in the fluence model, because the composition of the fuel materials varies with core exposure (i.e., burnup) and by cycle ( e.g., core loading).

3.2.2.3 Reactor Core Shroud The core shroud is a canister-like structure that surrounds the reactor core. It channels the reactor coolant and steam produced by the core into the steam separators. Axially the shroud extends

.almost the entire height of the model and is divided into three sections: lower, central, and upper.

The lower shroud extends from the bottom of the model to the core support plate flange, the central shroud extends from the core support plate flange to the top guide flange, and the upper shroud extends from top guide flange to the top of t~e shrou~ head rim.

The shroud structure is illustrated in Figure 3-3. It is shown that the lower shroud section is conical in design and is accurately modeled with conical elements. The central and upper shroud sections are cylindrical in design <:111d are. modeled with cylindrical elements.

The shroud structure is capped with a spherical-shaped shroud head that includes numerous penetrations for the steam separator standpipes that pass the steam from the core to the steam separators and dryers. The shroud head is semi-spherical in design and is modeled with spherical elements.. The steam separator standpipes that channel steam to the steam separators are cylindrical in design and are modeled with cylindrical elements.

3-16

I Neutron Fluence Calculation 3.2.2.4 Downcomer Region The downcomer region lies between the core shroud a1'd the RPV. The downcomer is effectively cylindrical in design, but with geometrical complexities created by the presence of jet pumps, surveillance capsules, and shroud repair tie rods. The majority of the downcomer region is modeled with cylindrical elements. The areas of the downcomet containing the jet pumps, surveillance capsules, and tie rods are modeled with the appropriate geometry elements to represent their.design features and to preserve their radial, azimuthal, and axial placement in the downcomer region. These structures are described further in the following subsections.

3.2.2.4.1 Jet Pumps Hatch Unit 2 has ten jet pump assemblies in the downcomer region, which provide the main recirculation flow for the core. In the fluence model, jet pump assemblies are positioned azimuthally at 30°;60° and 90°, which when symmetry is applied to the model represents all ten jet pump assemblies in the reactor. The 30° jet pump assembly represents the 30°, 150°, 210° and 330°jet pump assemblies; the 60° jet pump assembly represents the 60°, 120°, 240° and 300° jet pump assemblies; and the 90° jet pump assembly represents the 90° and 270° jet pump assemblies. Note there are no jet pump assemblies present at the 0° or 180° locations.

The jet pump assembly model includes representations for the riser, mixer, and diffuser pipes, nozzles, rams head, riser inlet pipe, and riser brace yoke, leaves arid pads. The jet pump assembly is modeled using cylindrical elements for the jet pump riser and mixer pipes. The mixer nozzles, adapters, and diffusers are modeled as stepped shells to represent the axially-varying radii. The riser pipe is correctly situated between the mixer pipes. Holding the riser pipe in place are riser brace elements, which are approximated as rectangular regions by radially segmented cylindrical arc*elements. The jet pump assembly includes hold down beams and brackets, built with rotating box elements that are attached to the rams head. Toroidal elements are used to model the rams head piping.

3.2.2.4.2 Surveillance Capsules Section 3.1 describes the surveillance capsules installed in the Hatch Unit 2 reactor. Over the reactor's operational history, there have been two types of surveillance capsules present: original equipment manufacturer (OEM) surveillance capsules and a reconstituted capsule.

The three (3) OEM surveillance capsules installed in the Hatch Unit 2 reactor are positioned in close proximity to the RPV inner wall surface atthe 30°, 120° and 3.00° azimuthal angles around the RPV. For this analysis, which assumes quadrant symmetry in the azimuthal dimension, the capsule modeled at30° represents the 30° capsule (OEM and reconstituted) and the.60° capsule represents the i20° and 300° capsules in the reactor. A separate flux wire holder is attached externally to the 30° surveillance capsule container. Although the flux wire holder is not shown in the figures of this report, it is ~ncluded in the model and is represented at the 30° capsule location.

The RPV surveillance capsules are rectangular in design but are approximated in the model with cylindrical elements to facilitate their inclusion in the cylindrical geometry that defines the downcomer region model. This model1ng approximation is acceptable due to the small view factor of the capsule relative to its radial 'distance from the reactor core. The coolant water that 3-17

Neutron Fluence Calculation surrounds the capsule containers on all sid(1s is explicitly modeled for its scattering and attenuation effects in the neutronics calculation.

3.2.2.4.3 Shroud Repair Tie Rods Four shroud repair tie rods were installed in the downcomer region of the Hatch Unit 2 reactor during refueling outage 12. It is noted that the tie rods were replaced during refueling outage 20.

The tie rods that are installed in the Hatch Unit 2 reactor are located circumferentially around the shroud at the 45°, 135°, 225° and 315° azimuths. The tie rods lie effectively in the spaces between the jet jump assemblies. All of the tie rods are represented by the 45° tie rod in the fluence model.

Three axial zones are used to represent the tie rods in the model: the lower mounting hardware, the tie rod shaft, and the upper mounting hardware. The tie rod shaft is cylindrical in design. In the azimuthal position, the tie rod lies vertically and centrally between the jet pump assemblies.

In the radial dimension, the top of the tie rod extends slightly outward from the bottom of the tie rod; therefore, showing a slight axial tilt relative to the vertical shroud wall. Due to the slight axial tilt, the shroud repair tie rods are modeled as a series of stacked cylinders in the downcomer region, with each cylinder representing a different radial distance from the shroud wall. The upper and lower hardware components are semi-homogenized and are composed of cylindrical components.

3.2.2.5 Reactor Pressure Vessel The RPV and RPV cladding lie outside the downcomer region, with both modeled using cylindrical elements. The cladding-to-base metal interface is a key location for RPV fluence calculations and is preserved in the model. This interface defines the inside surface (OT) for the RPV base metal where the RPV fluence is calculated. Hatch Unit 2 has cladding only on the inside surface oftheRPV wall. Representations of the forgings for the recirculation inlet (N2) and outlet (Nl) nozzles are included in the RPV wall. The nozzle forgings and safe-ends extend radially outward into the cavity region to the biological shield wall. The nozzle representations are modeled in their true cylindrical forms using cylindrical and conical elements to preserve their basic design features.

3.2.2.6 Thermal Insulation The RPV thermal insulation lies in the cavity region outside the RPV wall. The insulation is cylindrical in design and follows the contour of the RPV wall. It is modeled with cylindrical elements.

3.2.2.7 Inner and Outer Cavity There are effectively two cavity regions represented in the model. The inner cavity region lies between the outer surface of the RPV wall and the inner surface of the RPV insulation. The outer cavity region lies between.the outer surface of the RPV insulation and inner surface of the biological shield wall cladding. The cavity regions follow the cylindrical contours of the RPV wall, RPV insulation, and biological shield, and are therefore modeled with cylindrical elements.

3-18

I I I.

Neutron Fluence Calculation 3.2.2.8 Biological Shield Model The biological shield (concrete) defines the outermost region of the fluence model. The biological shield for the modeled elevations is effectively cylindrical in design and is modeled with cylindrical elements. Cladding is modeled on the inside and outside surfaces of the concrete wall.

3.2.2.9 Above-Core Components Figure 3-3 includes illustrations of other components and regions that lie above the reactor core region. The predominant above-core components represented in the model include the top guide, core spray sparger piping, upper core shroud wall, shroud head, and steam separator stand pipes.

The shroud regions and steam separator stand pipes are mentioned in further detail in Section 3.2.2.3.

3.2.2.9.1 Top Guide The top guide component lies above the core region. and is appropriately modeled to include discrete representations of the top guide plates and accounting for the fuel assembly parts, top guide pads, and coolant between the plates. The fuel assembly parts are modeled in two axial segments: the fuel rod plenum and the fuel assembly upper tie plate that includes the fuel rod upper end plugs. The top guide is modeled with combinations of rectangular and cylindrical elements. The top guide sits directly upon the shroud upper flange. The fuel assembly parts are modeled with the same elements as the reactor core region.

3.2.2.9.2 Core Spray Spargers and Piping The core spray spargers include upper and lower sparger.annulus pipes and a vertical inlet pipe.

The core spray spargers are appropriately represented as torus structures in the model. The sparger pipes reside inside the upper shroud wall above the top guide. The spargers are modeled as pipe-like structures and include a representation ofreactor coolant inside the pipes. The sparger spray nozzles are not represented individually. Instead, the region in which they reside is represented as a homogeneous cylindrical element of steam and stainless steel.

3.2.2.10 Below-Core Component Models Figure 3-3 includes illustrations of other components and regions that lie below the reactor core region. The predominant below core (i.e., below active fuel) components represented in the fluence model include the lower fuel assembly parts, fuel support pieces, core support plate, core support plate rim bolts, dry tubes, core support plate bypass plugs, cruciform control rods, control rod guide tubes, and lower shroud wall. The lower shroud wall and fuel assembly components are described in previous sections, with the remaining components described in the

. following subsections.

3.2.2.10.1 Lower Fuel Assembly Parts The lower fuel assembly parts that include the lower fuel assembly end plugs, tie plate and fuel nose piece for each unique fuel assembly design are appropriately represented in the fluence

,model-to provide flux and fluence attenuation of the neutral particles escaping the bottom of the 3-19

Neutron Fluence Calculation fuel and irradiating the below-core components. The lower fuel assembly parts are modeled with the same elements as the reactor core region.

3.2.2.10.2 Fuel Support Pieces The nuclear fuel assemblies loaded in the reactor are seated on fuel support pieces. Two types of fuel support pieces are included in the fluence model: a four-assembly fuel support piece, and a single-assembly fuel support piece for peripheral assembly locations. The four-assembly fuel support piece allows for the structure on which the fuel assembly is seated, the presence of a cruciform control rod and the associated coolant flow. Combinations of cylindrical, rectangular, and wedge elements are used to construct the fuel support piece models.

3.2.2.10.3 Core Support Plate, Rim Bolts, Dry Tubes, and Bypass Plugs The core support plate includes appropriate penetrations for the fuel support pieces, control rod guide tubes, cruciform control rods, dry tubes, core support plate bypass plugs, and the core support plate rim bolts. Core support plate rim bolts extend from a couple inches above the core support plate to below the shroud lower flange, traversing the core plate, core plate rim, and core shroud lower flange. The core support plate and rim are modeled using cylindrical and rectangular elements, while the rim bolts are modeled by cylindrical elements. The dry tubes and bypass plugs also extend above the core support plate and are cylindrical elements.

3.2.2.10.4 Control Blades and Guide Tubes The fluence model allows for the representation of cruciform-shaped control blades and tubular control blade guide tubes in *the below-core regions of the reactor. Coolant flow paths are included in the model to account for the scattering of neutrons in subcooled water conditions.

The control blade and guide tube components are modeled using combinations of rectangular and cylindrical elements.

3.2.2.11 Summary of the Geometry Modeling Approach To summarize, the reactor modeling process incorporates several key features that allow the reactor design to be accurately represented for RPV, RVI component, surveillance capsule, and other structural component fluence evaluations. Following is a summary of some of the key features of the model:

Combinations of rectangular, cylindrical, conical, spherical, toroidal, and wedge elements are used in the model to provide an accurate geometrical representations of reactor core, RPV, RVI components, support structures, and coolant regions in the reactor assembly.

The reactor core is modeled with rectangular elements to represent the true geometrical shape of the core. The fuel assemblies in the core region are also sub-meshed with additional rectangular and wedge elements to represent the power and isotopic distributions in the assemblies.

The fuel assembly tie plates, fuel rod end plugs, fuel nose piece, fuel channel and bypass flow regions are appropriately represented above and below the active fuel elevations. These components and regions are modeled using the same geometry elements as the reactor core.

3-20

Neutron Fluence Calculation A combination of rectangular and cylindrical elements is used to describe the transition parts between the rectangular boundary of the core region and the cylindrical boundary of the core shroud.

Cylindrical, conical, and wedge elements are used to model the components and regions that extend outward from the core region ( core shroud, downcomer, RPV, etc,).

Each jet pump assembly in the downcomer region includes appropriate representations for the riser p.ipe, mixer pipes, diffuser pipes, nozzles, couplers, rams head, hold down bracket, riser brace and yoke assembly, riser brace pads, etc. and is modeled using combinations of cylindrical, box, conical and wedge elements.

The RPV surveillance capsules are modeled with arc elements at their nominal radial, azimuthal, and elevational positions behind the jet pumps in the downcomer region. Using azimuthal symmetry conditions, all surveillance capsules are represented in the quadrant-symmetric fluence model.

The shroud repair tie rods are modeled with stacked cylindrical components and the hardware is modeled separately with cylindrical elements.

The RPV is modeled entirely with continuous cylindrical elements except for the elevations that contain recirculation inlet and outlet nozzles. The nozzles are composed of a combination of cylindrical and conical elements to preserve the true geometries of the nozzle forgings.

The above:..core region includes accurate and appropriate representations of the top guide, upper.fuel assembly parts, core spray sparger piping, upper core shroud, shroud head, and steam separator stand pipes. Combinations of rectangular parallelepiped, cylindrical, spherical, toroidal, and wedge elements are used to describe the above-core components, coolant flow regions, and bulk feedwater regions.

The below-core region includes appropriate representations for the lower fuel assembly parts, fuel support piece, core support plate, dry tubes, bypass plugs, core inlet regions, cruciform control rods, and control rod guide tubes. Combinations of,rectangular parallelepiped, cylindrical, spherical, toroidal, and wedge elements are used to describe the below-core components, coolant flow regions, and bulk inlet water regions.

The biological shield is appropriately represented as a cylinder with cladding on the inside and outside surfaces of the biological shield wall. The biological shield is described with cylindrical elements.

3.2.3Parametric Sensitivity Analyses Several plant-specific sensitivity analyses are performed to evaluate the accuracy and predictability of the neutral particle transport methodology for determining RPV, RPV Internals and surveillance capsule fluence in the Hatch Unit 2 reactor. Geometric meshing and numerical integration parameters are among the items pre-evaluated to ensure that the transport solution provides consistent results in all azimuthal, radial and axial dimensions of the reactor fluence model. The ultimate validation of the model is the demonstration that predicted activations for reactor surveillance dosimetry and boat samples, if available, meet the requirements of Regulatory Guide 1.190.

r 3-21 I

I

Neutron Fluence Calculation 3.2.4 Particle Transport Calculation Parameters The accuracy of the transport method is based on a numerical integration technique that employs ray-tracing to characterize the geometry, anisotropy treatments to determine the density and directional flow (i.e., angular flux) of particles, and convergence parameters to determine the overall accuracy of the converged flux. Plant-specific values are determined for each significant integration and ray-tracing parameter.

Two key parameters for the calculation of accurate angular flux are the angular quadrature set and Legendre order of scattering used in the transport calculation. The importance of the angular quadrature set is specifically addressed in Regulatory Position 1.3.5 of Regulatory Guide 1.190, where it is cited that an Ss angular quadrature (which is used in traditional transport models) may not be adequate when used in cavity transport calculations. The fluence model used in this analysis employs a higher-order S10 angular quadrature for all transport calculations to improve computational accuracy over the extended RPV beltline region.

The transport calculations also use the highest order of Legendre expansion of the scattering cross sections that is available on the nuclear data library for the anisotropy treatment. For the actinide and zirconium nuclides, this corresponds to a Ps expansion of the scattering cross sections, while for all other nuclides, a P7 expansion of the scattering cross sections is used.

Additional parameters of the calculation control the saturation of rays throughout the geometry by means of the parallel and axial ray densities as well as ray depths. The overall accuracy of the neutron flux calculation is determined using an iterative technique that converges the particle fluxes in all regions of the transport model.

3.2.5 Fission Spectrum and Neutron Source Modem core simulator software is capable of providing three-dimensional core power distributions and fuel isotopics in high-definition detail, viz., on a pin-by-pin basis. This allows fluence models to be constructed with a high-level of modeling detail for representing unique fission spectrum and neutron source terms for the transport calculation. This detail is incorporated into the fluence model.

The fission spectrum is determined for each transport calculation based on the relative weights of the contributing uranium and plutonium isotopes in the fuel materials. The fission spectra for the fuel actinides are derived from information that is provided in the BUGLE-96 nuclear data library [18] and the VITAMIN-B6 data library [19].

The spatial neutron source distribution is determined for each transport calculation using the pin-wise power density factors obtained from the core simulator software and data from the nuclear data library.

For activation and fluence evaluations, the peripheral fuel assemblies are specifically modeled to preserve the pin-wise power gradient at the core edge, as these bundles have the greatest effect on the determination of fluence in the RPV 3.3 Surveillance Capsule Activation and Fluence Results U.S. NRC Regulatory Guide 1.190 [15] requires that fluence calculational methods be validated by comparison to operating reactor dosimetry measurements. It is preferred that 3-22

Neutron Fluence Calculation measurement comparisons apply to the host reactor; however, provisiods are made that allow comparison to reactor dosimetry measurements of similar design. The acceptance criteria provided in Regulatory Guide 1.190 is that standard deviations determined from the calculated-to-measurement comparison ratios (C/M) fall within a computed standard deviation of+/- 20%.

This section presents the activation and fluence determined for two (2) sets of original equipment manufacturer (OEM) surveillance capsules that have been removed from the Hatch Unit 2 reactor. Computed activities are compared to the retrospective dosimetry measurements for the flux wires. It is determined that the total average C/M comparison ratio and standard deviation for the flux wire measurements is 1.05 +/- 0.13.

It is shown in Section 3.4 that the combined uncertainty for the 120° surveillance capsule is 10.5%. It is determined that the fluence methodology used to evaluate the Hatch Unit 2 reactor surveillance capsules meets the requirements of Regulatory Guide 1.190 with no discemable bias in the computed results. Therefore, these comparative results allow the computed best-estimate fluence calculated by the fluence methodology to be used without adjustment for the vessel fracture toughness analyses prescribed in U.S. NRC Regulatory Guide 1.99, Revision 2 [6].

The 300° OEM capsule and the 30° reconstituted capsule still remain in the reactor.

3.3.1 Summary of the Flux Wire Activation Analysis Two (2) sets of OEM flux wires have been removed from the Hatch Unit 2 reactor and retrospective dosimetry measurements performed. Activation and fast neutron fluence (E> 1.0 Me V) are determined for the following flux wire sets.'

Flux wires from the 30° flux wire holder removed at EOC 8 with an accumulated exposure of 6.61 EFPY [8], and Flux wires from the 120° capsule removed at EOC 24 with an accumulated exposure of 30.3 EFPY (See Appendix A).

3.3.1.1 Summary of the Surveillance Capsule Activation and Fluence Analysis Table 3-5 presents a summary of the average calculated-to-measured (C/M) results of specific activities and fluence for the OEM surveillance capsules and flux wire sets that were irradiated and removed from the Hatch Unit 2 reactor. A total of 15 flux wire measurements are evaluated.

The total average C/M and standard deviation for all measurement data is determined to be 1.09 +/- 0.11.

3-23

Neutron Fluence Calculation Table 3-5 Summary of Fluence and Activity Comparisons for the Hatch Unit 2 Dosimetry Time of Accumulated Fast Neutron Number of Calculated vs.

Standard Dosimeter Exposure Fluence (>1.0 Measured Removal (EFPY)

MeV, n/cm2)

Measurements (C/M)

Deviation (a) 30° Capsule EOC8 6.61 2.67E+17 9

1.04 0.08 120° Capsule EOC24 30.3 1.39E+18 6

1.16 0.11 Total Flux Wire 15 1.09 0.11 Average 3.3.2 Comparison of Predicted Activation to Plant-specific Measurements The comparison of predicted activations to measurements for the Hatch Unit 2 Cycle 8 and 24 flux wires are presented in this subsection. Fluence and lead factors for each capsule are reported in Subsection 3.3.3.

3.3.2.1 Cycle 8 30° Surveillance Capsule Activation Analysis Copper, iron, ~nd nickel flux wires were irradiated in the Hatch Unit 2 30° surveillance capsule during the first eight cycles of reactor operation. The wires were removed after being irradiated

\\

for a total of 6.61 EFPY [8]. Activation measurements were performed for the following reactions: 63Cu (n,a) 6°Co, 58Ni (n,p) 58Co, and 54Fe (n,p) 54Mn. The precise location of the individual wires within the capsule is not known; therefore, the activation calculations were performed at the center of the capsule container.

Table 3-6 provides a comparison of the calculated specific activities and the measured specific activities for each flux wire specimen. The average calculated-to-measured (C/M) ratio and standard deviation for the flux wires irradiated in the 30° capsule is determined to be 1.04 +/- 0.08.

3-24

Neutron Fluence Calculation Table 3-6 Comparison of Flux Wire Calculated-to-Measured Activities for the 30° Surveillance Capsule Removed from Hatch Unit 2 at EOC 8 Measured Calculated Calculated vs.

Standard Flux.Wires (dps/g)

(dps/g)

Measured (C/M)

Deviation (a)

Iron Iron 1 (65146) 9.03E+04 9.84E+04 1.09 Iron 2 (65147) 9.10E+04

.9.84E+04 1.08.

Iron 3 (65148) 9.24E+04 9.84E+04 1.07

, Iron Average 1.08 0.01 Nickel Nickel 1 (65146) 1.24E+06 1.39E+06 1.12 Nickel 2 (65147) 1.28E+06 1.39E+06 1.08 Nickel 3 (65148) 1.27E+06 1.39E+06 1.09 Nickel Average 1.10 0.02 Copper

\\

Copper 1 (65146) 1.13E+04 1.10E+04 0.97 Copper 2 (65147) 1.19E+04 1.10E+04 0.92 Copper 3 (65148) 1.21 E+04 1.10E+04 0.91 Copper Average 0.94 0.03 Total Flux Wire 1.04 0.08 Average 3.3.2.2 Cycle 24 120° Surveillance Capsule Activation Analysis

(

Copper, iron, and nickel flux wires were irradiated in the Hatch Unit 2 120° surveillance capsule during the first 24 cycles of reactor operation. The wires were removed after being irradiated for a total of 30.3 EFPY. Activation measurements were performed for the following reactions: 63Cu (n,a) 6°Co, 54Fe (n,p) 54Mn, and 58Ni (n,p) 58Co (See Appendix A). The precise location of the individual wires within the capsule is not known; therefore, the activation calculations were performed at the center of the capsule container.

Table 3-7 provides a comparison of the calculated specific activities and the measured specific activities for each flux wire specimen. The, average calculated-to-measured (C/M) ratio and standard deviation for the flux wires irradiated in the 120° capsule is determined to be 1.16

+/- 0.11.

3-25

Neutron Fluence palculation Table 3-7 Comparison of Flux Wire Calculated-to-Measured Activities for the 120° Surveillance Capsule Removed from Hatch Unit 2 at EOC 24 Flux Wires Measured Calculated Calculated vs.

Standard (dps/mg)

(dps/mg)

Measured (C/M)

Deviation (a)

Iron Iron G4 1.17E+02 1.33E+02 1.14 Iron GS 1.08E+02 1.33E+02 1.23 Iron Average 1.19 0.06 Nickel Nickel G4 1.43E+03 1.78E+03 1.2S Nickel GS 1.39E+03 1.78E+03 1.28 Nickel Average 1.26 0.02 Copper Copper G4 2.21 E+01 2.29E+01 1.04 Copper GS 2.20E+01 2.29E+01 1.04 Copper Average 1.04 0.01 Total Flux Wire 1.16 0.11 Average 3.3.3 Capsule Peak Fluence Calculations and Lead Factor Determinations Table 3-8 provides the best-estimate fast neutron fluence and lead factors for the surveillance capsules in the Hatch Unit 2 reactor. The fluence and lead factor for each capsule is reported at the time that the capsule was removed from the reactor.

Table 3-8 Best-Estimate Fluence and Lead Factors Determined for the Hatch Unit 2. Surveillance Capsu,es RPVOT RPV 1/4T Capsule Time of Capsule Fast Peak Removal Peak Fluence Lead Lead Fluence (n/cm2)

Fluence (n/cm2)

Factor 1 (n/cm2)

Factor 1 30° Capsule EOC8 2.67E+17 4.14E+17 0.64 2.91 E+17 0.92 120° Capsule EOC24 1.39E+18 1.96E+18 0.71 1.39E+18 1.00

1.

The lead factor is defined as the ratio of the fast neutron fluence at the center of the surveillance capsule to the peak fast neutron fluence at the base metal inner surface (OT) of the RPV. A second lead factor is provided assuming the peak fast neutron fluence at the 1 /4T depth of the RPV wall.

3-26 I

1

Neutron Fluence Calculation The average calculated-to-measured activation comparisons for the OEM surveillance capsules presented in the previous sections* show no discemable bias in the computational fluence method in accordance with U.S. Nuclear Regulatory Commission Regulatory Guide 1.190 [15].

Therefore, the best-estimate fluence reported for each capsule in Table 3-8 is the unbiased fast neutron fluence computed by the fluence methodology. This is discussed further in Section 3.4.

3.4 Capsule Fluence Uncertainty Analysis This section presents the combined uncertainty analysis and determination of bias for the Hatch Unit 2 OEM capsule fluence evaluation. The requirements for determining the combined uncertainty and bias for light water reactor fluence evaluations are proyided in Regulatory Guide 1.190. The combined uncertainty is comprised of two components: comparison uncertainty factors developed from plant measurements and analytic uncertainties. When combined, these components provide a basis for determining the overall uncertainty (lcr) and bias in the capsule fluence reported in this analysis; The method implemented for determining the combined uncertainty and bias for reactor component fluence is described in the RAMA Theory Manual [14]. Regarding the determination*

of a bias in the fluence, Regulatory Guide 1.190 provides that an adjustment to the calculated fluence for bias effects is needed if a statistically significant bias exists in the computational fluence method. It is determined in this report that the combined uncertainty (lcr) determined for the 120° capsule removed from the Hatch Unit 2 reactor at BOC 24 is +/- 10.5% and that no adjustment for bias effects is required to the calculated capsule fluence values presented in Section 3.3 of this report.

The following subsections describ_e the comparison uncertainties, the determination of the analytic uncertainty, and the det~rmination of the overall combined uncertainty and bias for the Hatch Unit 2 capsule fluence evaluation.

3.4. 1 Comparison Uncertainty Comparison uncertainty factors are determined by comparing calculated activities with activity measurements. For capsule fluence evaluations, two comparison uncertainty factors are considered: plant-specific comparison factors and benchmark comparison factors. Comparison uncertainty factors based upon measurements also involve the combination of two components:

the calculated-to-measured (C/M) activity ratio and a measurement uncertainty.

3.4.1.1 Plant-Specific Comparison Uncertainty The Hatch Unit 2 reactor is a BWR/4 class plant with a core loading of 560 fuel assemblies. A total of 192 plant-specific dosimetry measurements have been evaluated for BWR/4 class plants of various core configurations using the RAMA Fluence Methodology. The evaluations resulted in an overall unbiased C/M and standard deviation of 1.02 +/- 0.11. This result is used to determine the Hatch Unit 2 uncertainty factor.

3.4.1.2 Benchmark Comparison Uncertainty The benchmark/simulator comparison uncertainty used in the Hatch Unit 2 uncertainty analysis is based on a set of industry standard simulation benchmark comparisons. In accordance with 3-27

Neutron Fluence Calculation Regulatory Guide 1.190, it is appropriate to include comparisons of vessel simulation benchmark measurements in the overall uncertainty evaluation. Specifically, the VENUS-3 and PCA benchmarking results [20] are included in the comparative analysis.

3.4.2 Analytic Uncertainty The computational models used for fluence analyses are comprised of numerous analytical parameters that have associated uncertainties in their values. The uncertainty in these parameters are tested for their contribution to the overall fluence uncertainty.

The uncertainty values for the geometry parameters are based upon tolerances in the nominal dimensional data and standard deviations in measured dimensions used to construct the plant geometry model. The uncertainty values for the material parameters are based upon tolerances in the material densities for the water and nuclear fuel materials and the compositional makeup of typical steel materials.

The uncertainty values for the fission source parameters are based upon uncertainties in the fuel exposure and power factors for the fuel assemblies loaded on the core periphery. The transport method used in the fluence analysis employs a fission source calculatioµ that accounts for the relative contributions of the uranium and plutonium fissile isotopes in the fuel and the relative power density of the fuel in the reactor. Both fission source parameters are derived directly from information calculated by three-dimensional core simulator codes. The uncertainty values for the nuclear cross section parameters are based upon uncertainties in the number densities for the predominant nuclides that make up the reactor materials.

The uncertainty parameters for the computational fluence method are based upon geometry meshing and the numerical integration parameters used in the neutronics transport calculation.

Several meshing and numerical integration parameter sensitivity cases are run to determine the ideal values for the transport calculation and the resulting impact on the analytic uncertainty.

3.4.3 Combined Uncertainty The combined uncertainty for a capsule fluence evaluation is determined with a weighting function that combines the analytic, plant-specific comparison, and benchmark comparison uncertainty factors. Table 3-9 shows the combined bias and uncertainty (lo) factors determined for each RPV surveillance capsule from the Hatch Unit 2 reactor. It is shown that the combined uncertainty in the fast neutron fluence (E > 1.0 MeV) determined for the Hatch Unit 2 120° capsule removed at EOC 24 is 10.5%, and that no significant bias exists in the computed results.

Therefore, no bias correction is applied to the capsule fluence values reported in Section 3.3 of this report.

3-28

Neutron Fluence Calculation Table 3-9 Hatch Unit 2 Surveillance Capsule Combined Uncertainty for Energy >1.0 MeV Time of Accumulated Combined Capsule Removal Exposure Combined Bias1 Uncertainty (1a)

(EFPY) 30° EOC8 6.61 0.0%

10.5%

120° EOC24 30.3 0.0%

10.5%

1. The bias terms are less than their constituent uncertainty values, concluding that no statistically-significant bias exists.

3-29

4 CHARPY TEST DATA 4.1 Charpy Test Procedure Charpy impact tests were conducted in accordance with ASTM Standards El85-82 [3] and E23-02 [9]. The 1982 version of E185 has been reviewed and approved by NRC for surveillance capsule testing applications. This standard references ASTM E23. The tests were conducted using a Tinius Olsen Testing Machine Company, Inc. Model 84 impact test machine with a 300 ft-lb ( 406.75 J) energy capacity. The Model 84 is equipped with a dial gage as well as the MPM optical encoder system for accurate absorbed energy measurement. The machine is also equipped with an instrumented striker, so a total of three independent measurements of the absorbed energy were made for every test. In all cases, the optical encoder measured energy was reported as the impact energy. The optical encoder energy is much more accurate than the analog dial. The optical encoder can resolve the energy to within 0.04 ft-lbs (0.054 J), whereas, for the dial, the resolution is around 0.25 ft-lbs (0.34 J). The impact energy was corrected for windage and friction for each test performed. The velocity of the striker at impact was nominally 18 ft/s (5.49 mis). The MPM encoder system measures the exact impact velocity for every test. Calibration of the machine was verified as specified in ASTM E23, and verification specimens were obtained from the National Institute for Standards and Technology (NIST) and tested in accordance with the standard.

The ASTM E23 procedure for specimen temperature control using an in-situ heating and cooling system was followed. The advantage of using the MPM in-situ heating/cooling technology is that each specimen is thermally conditioned right up to the instant of impact. Thermal losses associated with liquid bath systems, such as those resulting from transfer of a specimen from a liquid bath to the test machine, are completely eliminated. Each specimen was held at the desired test temperature for at least 5 minutes prior to testing, and the fracture process zone temperature was held to within+/- 1.8 F (+/- 1 C) up to the instant of strike. Precision calibrated tongs were used for specimen centering on the test machine.

Lateral expansion (LE) was determined from measurements made with a lateral expansion gage.

The lateral expansion gage was calibrated using precision gage blocks which are traceable to NIST. The percentage of shear fracture area was determined by integrating the ductile and brittle fracture areas using the MPM Digital Optical Comparator (DOC) image analysis system. As shown in Figure 4-1, each fracture surface image is captured, outlined to delineate the brittle area, and outlined to define the outer ductile fracture region. The DOC software then performs a pixel area integration and automatically calculates the shear fracture area. This method for shear area determination is the most accurate method given in ASTM E23, and is far superior to the commonly used photograph comparison method.

4-1

Charpy Test Data The number of Charpy specimens for measurement of the transition region and upper shelf was limited. Therefore, the choice of test temperatures was very important. Prior to testing, the Charpy energy-temperature curve was predicted using embrittlement models and previous data.

The first test was then conducted near the middle of the transition region, and test temperature decisions were then made based on the test results. Overall, the goal was to perform two or three tests on the upper shelf, and to use the remaining specimens to characterize the 30 ft-lb (41 J) index. This approach was successful and the transition region and upper shelf energy are well defined.

Figure 4-1 Illustration of Digital Optical Comparator Measurement of Shear Fracture Area First, the Brittle Fracture Area is Outlined (within green line). Next, the Outer Ductile Fracture Area is Outlined (within red line). Finally, the Software Integrates the Areas and Calculates the Percent Shear Fracture Area.

4.2 Charpy Test Data for the 120° Capsule A total of eight irradiated base, eight irradiated weld, and eight irradiated HAZ metal specimens were tested over the transition region temperature range and on the upper shelf. The data are summarized in Tables 4-1 through 4-3. In addition to the energy absorbed by the specimen during impact, the measured lateral expansion values and the percentage shear fracture area for each test specimen are listed in the tables. The Charpy energy was acquired from the optical encoder signal and has been corrected for windage and friction in accordance with ASTM E23.

The impact energy is the energy required to initiate and propagate a crack in the Charpy 4-2

Charpy Test Data specimen. The optical encoder and the dial cannot correct for tossing energy or losses in the test machine, and therefore this small amount of additional energy, if present, may be included in the data for some tests. The instrumented striker energy does not include tossing energy or machine vibration energy since the energy, in this case, is measured only during a few milliseconds of contact between the striker and specimen. Based ori comparison between the instrumented striker energy and the optical encoder energy, it has been shown that the tossing energy, and other losses, are small for most tests.

The lateral expansion is a mea~ure of the transverse plastic deformation produced by the contact edge of the striker during the impact event. Lateral expansion is determined by measuring the maximum change of specimen thickness along the sides of the specimen. Lateral expansion is a measure of the ductility of the specimen. The nuclear industry tracks the embrittlement shift using the 35 mil (0.89 mm) lateral expansion index. In accordance with ASTM E23, the lateral expansion for some specimens, which could not be broken by hand after the impact test, should not be reported as broken since the lateral expansion of the unbroken specimen is less than that for the broken specimen. Therefore, when these conditions exist, the value listed is the unbroken measurement and a footnote is included to identify these specimens. All of the 120° capsule specimens that did not separate during the test could be broken by hand under the ASTM E23 requirements.

The percentage of shear fracture area is a direct quantification of the transition in the fracture modes as the temperature increases. All metals with a body centered cubic lattice structure, such, as ferritic pressure vessel materials, undergo a transition in fracture modes. At low test temperatures, a crack propagates in a brittle manner and cleaves across the grains. As the temperature increases, the percentage of shear ( or ductile) fracture increases. This temperature range is referred to as the transit10n region and the fracture process is mixed mode. As the temperature increases further, the fracture process is eventually completely ductile (i.e., no brittle component) and this temperature range is referred to as the upper shelf region.

Table 4-1 Irradiated Charpy V-Notch Impact Test Results for Surveillance Base Metal Specimens (Heat C8554) from the Hatch Unit 2 120° Surveillance Capsule Base Irradiated: Heat C8554, Longitudinal, 120° Capsule Test Lateral Specimen Temperature Impact Energy Expansioi;i Percent Shear ID OF

(°C) ft-lb (J) mils (mm)

P1-16

-102.1

(-74.5) 17.05 (23.12) 10.0 (0.25) 6.4 P1-18

-44.0

(-42.2) 17.95 (24:34) 14.5 (0.37) 18.1 P1-17

-1.8

(-18.8) 34.34 (46.56) 25.6 (0.65) 24.4 P1-19 34.9 (1.6) 49.74 (67.44) 39.8 (1.01) 50.6 P1-13 72;0 (22.2)

73.62 (99.81) 53.7 (1.36) 70.3 P1-20 111.4 (44.1) 110.37 (149.64) 76.0 (1.93) 91.3 P1-14 152.8 (67.1) 112.70 (152.80) 72.3 (1.84) 1do.o P1-15 374.0 (190.0) 111.64 (151.36) 73.2 (1.86) 100.0 4-3

Charpy Test Data Table 4-2 Irradiated Charpy V-Notch Impact Test Results for Surveillance Weld Metal Specimens (Heat 51912) from the Hatch Unit 2 120° Survemance Capsule Weld Irradiated: Heat 51912, 120° Capsule Test Lateral Specimen Temperature Impact Energy Expansion Percent Shear ID Of

, (OC) ft-lb (J) mils (mm)

P2-12

-103.2

(-75.1) 4.03 (5.46) 0.9 (0.02) 3.2 P2-22

..:so.a

(-45.6) 8.77 (11.89) 7.2 (0.18) 14.8 P2-24

-21.5

(-29.7) 35.51 (48.14) 24.7 (0.63) 24.8.

P2-21

-1.5

(-18.6) 60.82 (82.46) 41:9 (1.06) 43.6 P2-23 36.5 (2.5) 72.13 (97.79) 46.3

( 1.18) 57.3 P2~9 71.6 (22.0).

98.17 (133.10) 64.8

({65) 78.4 P2-10 172.0 (77.8) 151.24 (205.05) 84.2 (2.14) 100.0 P2-11 385.3 (196.3) 124.22 (168.42) 75.3 (1.91) 100.0 Table4-3

Irradiated Charpy V-Notch Impact Test Results for Surveillance HAZ Metal Specimens from the Hatch Unit 2 120° Surveillance Capsule HAZ Irradiated: 120° Capsule Test Lateral Specimen Temperature Impact Energy Expansion Percent Shear ID Of (OC) ft-lb (J) mils (mm)

P3-4

-101.4

(-74.1) 11.72 (15.89) 9.2 (0.23) 9.6 P3-6

-51.7

(-46.5) 35.31 (47.87) 24.0 (0.61) 24.1 P3-5 1.0

(-17.2) 32.02 (43.41) 26.5 (0.67) 31.9 P3-7 33.8..

(1.0) 66.04 (89.54) 45.2 (1.15) 50.9 P3-1 72.0 (22.2).

98.62 (133.71) 66.0 (1.68) 64.3 P3-8 112.3 (44.6) 159.27 (215.94) 78.2 (1.99) 100.0 P3-2 162.0 (72.2) 136.19 (184.65) 79.9 (2.03) 100.0 P3-3 382.1 (194.5) 127.53 (172.91) 73.5 (1.87) 100.0 4-4

5 CHARPY TEST RESULTS 5.1 Analysis of Impact Test Results For analysis of the Charpy test data, the BWRVIP ISP has selected the hyperbolic tangent (tanh) function as the statistical curve-fit tool to model the transition temperature toughness data. A hyperbolic tangent curve-fitting program named CVGRAPH [11] was used to fit the Charpy V-notch (CVN) energy and lateral expansion data. Analysis methodology (e.g., definition of upper fixed shelf and lower shelf) followed the BWRVIP conventions established for analysis of all ISP data [21]. The impact energy curve-fit from CVGRAPHare provided in Figure 5-1 (plate heat C8554) and Figure 5-2 (weld heat 51912). The lateral expansion curve fits ate provided in Figure 5-3 (plate heat C8554) and Figure 5-4 (weld heat 51912). HAZ results are not used in the BWRVIP ISP; thus, the HAZ data were not fit.

For the analysis of Charpy energy test data, lower shelf energy was fixed at 2.5 ft-lbs (3.4 J).

Upper shelf energy was foced at the average of all test energies exhibiting shear greater than or equal to 95%, consistent with ASTM Standard E185-82 [3]. For analysis of the lateral expansion test data, the lower shelf was fixed at 1.0 mils; the fixed upper shelf was defined as the average of the lateral expansion test data points exhibiting shear greater than or equal to 95%, consistent with the approach used for upper shelf energy.

5.2 Irradiated Versus Unirradiated CVN Properties Table 5-1 summarizes the T30 [30 ft-lb (41 J) Transition Temperature], T3smil [35 mil (0.89 mm)

'Lateral Expansion Temperature], Tso [50 ft-lb (68 J) Transition Temperature], and Upper Shelf Energy for the unirradiated and irradiated materials and shows the change (shift) from baseline values. The unirradiated values ofT30 and Tso were taken from the CVGRAPH fits provided in Figures 2-5 and 2-6; the unirradiated values. ofLsmiJ were taken from the CVGRAPH fits provided in Figures 2-7 and 2-8. The irradiated values are from the index temperatures determined in Figures 5-1 and 5-2 for impact energy and Figures 5-3 and 5-4 for lateral expans10n.

Table 5-2 provides a comparison of the measured T30 shift to the predicted shift for plate heat C8554 and weld heat 51912. Predicted shift is based on the formula provided in Regulatory Positio~ 1.1 of Reg. Guide 1.99, Rev. 2 [6] as shown in Note 3 to Table 5-2. The fluence was input as 13.9 x 1017 n/cm2, as reported in Table 3-8 for the 120° surveillance capsule. The measured shift for the surveillance plate and weld are less than the value expected ( e.g., the

measured shift is less than predicted shift+ margin).

Measured percent decrease in USE is presented in Table 5-3 and compared to the percent decrease predicted by Regulatory Position 1.2 and Figure 2 of Reg. Guide 1.99, Rev. 2. The measured percent decrease in USE for the surveillance plate and weld are less than the predicted percent decrease.

5-1

Charpy Test Results IRRADIATED PLATE HEAT C8554 (HA2-120)

CVGraph 6.02: Hyperbolic Tangent Curve Printed on 4/20/2018 11 :27 AM A= 57.34 B = 54.84 C = 77.46 TO= 37.47 D = 0.00 Correlation Coefficient= 0.986 Equation is A +B * [Tanh((T-TO)/{C+DT))]

Upper ShelfEnexgy = 112.17 (Fixed)

Lower Shelf Energy= 2.50 (Fixed)

Temp@30 ft-lbs= -4.90° F Temp@35 ft-lbs= 4.00° F Tenip@50 ft-lbs= 27.10° F Plant: Hatch 2 Orientation: LT Cll

,.Q -

I bJ).

~

Cl.) =

~-z

> u 100 40 Material: SA533B1 Capsule: 120 DEG Heat: C8554 Fluence: n/a

-200

-100 0

100 200 300 400 500 600 Temperature (° F)

CVGraph 6.02 04/20/2018 Page 1/2 Figure 5-1.

Irradiated Plate Heat C8554 Charpy Energy Plot (Hatch Unit 2 120° Capsule) (LT) 5-2

Plant: Hatch 2 Orientation: LT Material: SA533Bl Capsule: llO DEG Charpy Test Results Heat: C8554 Fluence: n/a IRRADIATED PLATE HEAT C8554 (HA2-120)

Charpy V-N otcb Data Temperature (0 F)

InputCVN Computed CVN Differential

-102 17:l 5.4 l'l.64

-44 18Xi 14.4 3.52

-2 34:3 31.7 2.64 35 49.7 55.5

-5.78 72 73.6 80.3

-6.66 Ill 110.4 98.0 12.36 153 112.7 106:9 5.84 374 111.6 112.2

-0.51 CVGraph 6.02 04/20/2018 Page 2/2 Figure 5_-1 (Continued)

Irradiated Plate Heat C8554 Charpy Energy Plot (Hatch Unit 2 120° Capsule) (LT)

Charpy Test Results IRRADIATED WELD HEAT 51912 (HA2-120)

CVGraph 6.02: Hypeibolic Tangent Curve Printed on 6/25/2018 I :53 PM A= 70.11 B = 67.61 C = 79.93 TO= 27.19 D = 0.00 Correlation Coefficient= 0.978 Equation is A+ B * [Tanh((T-TO)/(C+D1))]

Upper Shelf Energy= 137. 73 (Fixed)

Lower Shelf Energy= 2.50 (Fixed)

Temp@30 ft-lbs=-27.30° F Temp@35 ft-lbs=-18.80° F Temp@50 ft-lbs= 2.70° F Plant Hatch 2 Orientation: NA Material: SAW Capsule: 120 DEG Heat 51912 Fluence: n/a 160,--~~.--~~,--~~,--~~....... ~~....-~. ~....-~~-,--~.~..,....~----.

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-300

-200

-100 0

100 200 300 400 500 600 Temperature {° F)

CVGraph 6.02 06/25/2018 Page 1/2 Figure 5-2 Irradiated Weld Heat 51912 Charpy Energy Plot (Hatch Unit 2 120° Capsule) 5-4

Plant: Hatch 2 Orientation: NA Material: SAW Capsule: 120 DEG Charpy Test Results Heat:.51912 Fluence: n/a IRRADIATED WELD HEAT 51912 (HA2-120)

Charpy V-Notch Data Temperature {° F)

InputCVN Computed CVN Differential

-103 4.0 7.5

-3.46

-50 8.8 19.6

-10.85

-22 35.5 33.4 2.15,

-2 60.8 46:8 13.99 37 72:l 78,0

-5.82 72 98.2 104.2

-6.07 172 151.2 134.2 17.03 385 124.2 137.7

-13.49 CVGraph 6.02 06/25/2018 Page 2/2 Figure 5-2 (Continued)

Irradiated Plate Heat 51912 Charpy Energy Plot (Hatch Unit 2 120° Capsule) 5-5

Charpy Test Results 5-6 IRRADIATED PLATE HEAT C8554 LE (HA2-120)

CVGraph 6.02: Hyperoolic Tangent Curve Printed on 4/20/2018 11 :30 AM A= 36.90 B = 35.90 C = 78.47 TO= 22.12 D = 0.00 Correlation Coefficient= 0.986 Equation is A+ B * [Tanh((T-TO)/(C+D1))]

Upper ShelfL.E. = 72.80 (Fixed)

Lower ShelfL.E. = 1.00 (Fixed)

Temp@35 mils= 18.00° F Plant Hatch 2 Orientation: LT Material: SA533Bl Capsule: 120 DEG Heat: C8554 Fluence: n/a 80.-----.---...... ---.---...... ----.----,-----, ---...-~.---~----.---..,....---.---...,..---~--,------,---,

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-300

-200

-100 0

100 200 300 400 500 600 Temperature (0 F)

CVGraph 6.02 04/20/2018 Page 1/2 Figure 5-3.

Irradiated Plate Heat C8554 Lateral Expansion Plot (Hatch Unit 2 120° Capsule) (LT)

Charpy Test Results

. Plant: Hatch 2

. Orientation: LT Material: SA533Bl Capsule: 120 DEG IRRADIATED PLATE HEAT C8554 LE (HA2-120)

Charpy V-N otch Data Heat' C85.54 Fluence: n/a Temperature {° F) inputL.R Computed L. E.

Differential

-102 10.0 3.9 6.09

-44 14.5 12:2 2.27

-2 25.6 26.3

-0.68 35 39.8 42.7

-2.90 72 53.7 57.1

-3.37 111 76.0 66]

9.89 153 72.3 70.3 1.98 374 73:2 72:8 0.41 CVGraph 6.02 04/20/2018 Page 2/2 Figure 5-3 (Continued)

Irradiated Plate Heat C8554 Lateral Expansion Plot (Hatch Unit 2 120° Capsule) (LT) 5-7

Charpy Test Results IRRADIATED WELD HEAT 51912 LE (HA2-120)

CVGraph 6.02: Hyperbolic Tangent Curve Printed on 6/25/2018 3:37 PM A= 40.40 B = 39.40 C = 75.41 TO = 13.12 D = 0.00 Correlation Coefficient= 0.984 Equation is A+ B * [Ianh((T-TO)/(C+DT))]

Upper ShelfL.E. = 79.80 (Fixed)

Lower ShelfL.E. = 1.00 (Fi.xed)

Temp@35 mils= 2.80° F Plant: Hatch 2 Orientation: NA Material: SAW Capsnle: 120 DEG Heat 51912 Fluence: n/a 90.-----------------------------.---------------------------....... ----------------------------.

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-300

-200

-100 0

100 200 300 400 500 600 Temperature {° F)

CVGraph 6.02 06/25/2018 Page 1/2 Figure 5-4 Irradiated Weld Heat 51912 Lateral Expansion Plot (Hatch Unit 2120° Capsule) 5-8

Plant: Hatch 2 Material: SAW Orientation: NA Capsule: UO DEG Charpy Test Results Heat: 519U Fluence: n/a IRRADIATED WELD HEAT 51912 LE (HA2-120)

Charpy V-Notch Data Temperature (0 F) inputL. E.

Computed L. E.

DHTerential

-103 0.9 4.4

-3.55

-50 7.2 13.4

-6.24

-22 24;7 23.5 1.21

-2 41.9 32,9 9.04 37 46.3 52.2

-5.94 72 64.8 66.0

-1.21 172 84,2 78.7 5.55 385 753 79.8

-4.50 CVGraph 6.02 06/25/2018 Page 2/2 Figure 5-4 (Continued)

Irradiated Weld He~t 51912 Lateral Expansion Plot (Hatch Unit 2 120° Capsule) 5-9

Charpy Test Results Table 5-1 Effect of Irradiation {E>1.0 MeV) on the Notch Toughness Properties T 30* 30 ft-lb (40. 7 J)

Tso, 50 ft-lb (67.8 J)

T3smil135 mil CVN-Upper Shelf Energy Transition Temperature Transition Temperature (0.89 mm) Lateral (USE)

Material Expansion Temperature Identity Unirrad Irradiated AT30 Unirrad Irradiated AT50*

Unirrad Irradiated AT3smil Unirrad Irradiated Change OF OF OF OF OF OF OF OF OF ft-lb

<<~lb ft-lb (OC)

(OC)

(J)

C8554

-19.6

-4.9 14.7 6.2' 27.1 20.9 8.1 18.0 9.9 111.5 112.2 0.7 (LT orientation)

(-28.7)

(-20.5)

(8.2)

(-14.3)

(-2.7)

( 11.6)

(-13.3)

(-7.8)

(5.5)

(151.2)

. (152.1)

(0.9) 51912

-21.0

-27.3

-6.3 4.0 2.7

-1.3 7.3 2.8

-4.5 120.8 137. 7 16.9

(-29.4)

(-32.9)

(-3.5)

(-15.6)

(-16.3)

(-0.7)

(-13.7)

(-16.2)

(-2.5)

(163.8)

(186.8)

(23.0) 5-10

Charpy Test Results Table 5-2 Comparison of Actual Versus Predicted Emb~ittlement Fluence RG 1.99 Rev. 2 RG 1.99 Rev. 2 Identity Material (E>1.0 MeV, x1017 Measured Shift2 Predicted Shift3 Predicted OF (OC)

Shift+Margin3*4 n/cm2)1 OF (OC)

OF (OC)

C8554 Hatch Unit 2 surveillance plate 2.67 14.7 (8.2) 24.8 (13.8) 49.6 (27.5)

(LT orientation) 51912 Hatch Unit 2 surveillance weld 13.9

-6.3 (-3.5) 32.6 (18.1) 65.1 (36.2)

1.

Fluence value is reported in Table 3-8.

2.

The measured shift is taken from Table 5-1. -

3.

Predicted shift= CF x FF, where CF is a Chemistry Factor taken from the base metal table in USN RC RG 1.99, Rev. 2 [6], based on each material's Cu/Ni content, and FF is Fluence Factor, t<J.28-0.10 log f, where f = fluence in units of 1019 n/cm2 (E > 1.0 MeV) specified.

4.

Margin = 2'1(o;2 + ot.2}, where o; = the standard deviation on initial RT NDT (o; is taken to be 0°F), and 01',, is the standard deviation on b.RT NDT (28°F for welds and 17°F for base materials, except that O!',, need not exceed 0.50 times the mean value of b.RTNor). Thus, margin is defined as 34°F for plate materials and 56°F for weld materials, or margin equals shil'.t (whichever is less), per Reg. Guide 1.99, Rev. 2.

Table 5-3 Percent Decrease in Upper Shelf Energy Fluence Measured Predicted Decrease Identity Material (E>1.0 MeV, x1017 Decrease in USE in USE2 n/cm2)1

(%).

(%)

C8554 Hatch Unit 2 surveillance plate 2.67

__ 3 10.7 (LT orientation) 51912 Hatch Unit 2 surveillance weld 13.9

__ 3 16.9

1.

Fluence value is reported in Table 3-8.

2.

Based on the equations for Figure 2 of Reg. Guide 1.99, Rev. 2 [6] as provided in Reg. Guide 1.162 [22].

3.

V?lue less than zero 5-11

6 REFERENCES

1. 10 CFR 50, Appendices G (Fracture Toughness Requirements) and H (Reactor Vessel Material Surveillance Program Requirements), Federal Register, Volume 60, No. 243, dated December 19, 1995.

2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, "Rules for In service Inspection of Nuclear Power Plant Components,"

Nonmandatory_Appendix G, Fracture Toughness Criteria for Protection Against Failure.

3. ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), American Society for Testing and Materials, Philadelphia, PA, 1982.

4. BWRVIP-86, Revision 1-A: BWR Vessel and InternalsProject, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012. 1025144.

5. 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," August 28, 2007.

6. U.S. NRC Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"

Revision 2, May 1988.

7. "Guideline for the Management of Materials Issues," NEI 03-08, Nuclear Energy Institute, Washington, DC, Latest Edition.

8. GE Nuclear Energy, "E. I. Hatch Nuclear Power Station, Unit 2 Vessel Surveillance Materials Testing and Fracture Toughness Analysis," SASR 90-104, DRF Bl 1-00495, May, 1991.

9. ASTM Standard E23, Standard Test Methods fOr Notch Bar Impact Testing of Metallic Materials, ASTM International, West Conshohocken, PA, www.astm.6tg.

10. Material Test Certificate, Plate Heat C8554, Lukens Steel Company, File No. 1771-02-02, Job No. A98539, 1-27-71.

11. CV GRAPH, Hyperbolic Tangent Curve Fitting Program, Developed by ATI Consulting, Version 6.02, April 2014.

12: BWRVIP-126; Revision 2: BWR Vessel Internals Project, RAMA Fluence Methodology Software, Version 1.20. EPRI, Palo Alto, CA: 2010; 1020240.

13. BWRVIP-12I'-A: BWR Vessf!l and Internals Project, RAMA Fluence Methodology Procedures. Manual, EPRI, Palo Alto, CA: 2009. 1019052.

14. B WR VIP-114-A: B WR Vessel and Internals Project, RAMA Fluence Methodology Theory Manual, EPRI, Palo Alto, CA: 2009. 1019049.

6-1

References

15. U.S. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

16. "Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station," Docket Number 50-443, dated June 2012.

17. J. R. Askew, "A Characteristics Formulation of the Neutron Transport Equation in Complicated Geometries," United Kingdom Atomic Energy Authority, AEEW-M 1108, 1972.

18. "BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived

from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," RSICC Data Library Collection, ~L2C-185, March 1996.

19. "VITAMIN-B6: A Fine-Group Cross Section Library B_ased on ENDF/B-VI Release 3 for Radiation Transport Applications," RSICC Data Library Collection, DLC-184, December 1996.

20. BWR.VIP-115-A: BWR Vessel and Internals Project, RAMA Fluence Methodology Benchmark Manual - Evaluation of Regulatory Guide 1.190 Benchmark Problems, EPRI, Palo Alto, CA: 2009. 1019050.

21. BWRVIP-135, Revision 3: BWR. Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014.

3002003144.

22. U.S. NRC Regulatory Guide 1.162, "Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels," February 1996.

6-2

A DOSIMETER ANALYSIS A.1 Dosimeter Material Description The Hatch Unit 2 120° surveillance capsule dosimeter materials are pure Inetal wire~ which were located within the surveillance capsule along the ends of the Charpy specimens. The wire types provided for the Hatch Unit 2 surveillance program are iron; nickel, and copper. Each wire js nominally three inches (7. 62 cm) long. Further discussion of the dosimeter cleaning and mass measurements follows.

A.2 Dosimeter Cleaning and Mass Measurement At the time the surveillance capsule Charpy packets were opened, the dosimeter wires were cleaned in an ultrasonic cleaner in an acetone bath and were wiped with acetone wetted wipes to remove loose contamination. Upon receipt at the radiometric laboratory, the wires were visually inspected under a low magnification optical microscope. There was evidence of oxidation indicating the need for chemical etching and further cleaning. This was accomplished by soaking the Fe wire segments in a 4N solution of hydrochloric acid until the oxidation was etched from the surface. Similarly, the Cu and Ni wires were immersed in a 2N solution of nitric acid. The wires were then rinsed with distilled water, wiped once more with ethanol, and then

allowed to dry in air at room temperature. The wires then exhibited a clean, shiny appearance.

Figures A-1 through A-6 show low-power magnifications of the dosimetry wires as they were fo11nd prior. to cleaning, and after cleaning and coiling.

The total mass of each wire was measured using a Mettler Toledo XS 105DU analytical digital balance. Table A-I lists the results of these measurements, as well as the identification assigned to each dosimeter. The dosimeter identifications were assigned as the packet ID containing the dosimeter wire and type of dosimeter material.

As previously mentioned, the wires were tightly coiled for subsequent countirig and weighing.

Each wire was wrapped around a thin metal rod to form a coil of approximately 0.5 inch (12.7 mm) diameter or less, which yields a good approximation to a point source geometry at the distance the dosimeter wires are placed from the gamma detector. The coiled wire segments were pressed firmly against a hard surface to flatten the coil to yield the best counting geometry.

A.3 Radiometric Analysis.

Radiometric analysis was performed using high resolution gamma emission spectroscopy. In this* method, gamma emissions from the dosimeter materials are detected and quantified using solid-st.ate gamma ray detectors and computer-based signal processing and spectrum analysis.

The specifications of the gamma ray spectrometer system (GRSS) are listed in Table A-2. The GRSS features a hyper pure germanium (HPGe) detector that is housed in a lead-copper shield to reduce background count rates. Standard background subtraction procedures were used.

A-1

Dosimeter Analysis GRSS calibration was performed using a National Institute for Standards and Technology (NIST) traceable mixed gamma quasi-point source. The Canberra analysis software provides the capability for energy resolution and efficiency calibration using specified standard source information. Calibration information is stored on magnetic disk for U:se by the spectrographic analysis software package.

Since detector efficiency depends on the source-detector geometry, a fixed-reproducible geometry must be selected for the gamma spectrographic analysjs of the dosimeter materials.

For the dosimeter wires, the counting geometry was that of a quasi-point source. ( coiled wire) placed five inches (12.7 cm) vertically from the top surface of the detector shell.. Ii1 this way, extended sources up to 0.5 inch (1.27 cm) can be analyzed with a good approximation to a point source. The coiled. wires were well within the area needed to approximate a point source geometry. The HPGe detector was calibrated for efficiency using the NIST traceable source.

The accuracy of the efficiency calibration was checked using a gamma spectrographic analysis of the NIST traceable mixed gamma source. The isotopes contained in the source emit gamma rays which span the energy response of the detector for the dosimeter materials. These measurements show that the efficiency calibration is providing a valid measurement of source activity. The acceptance criteria for these measurements are that the software must yield a valid isotopic identification, and that the quantified activity of each correctly identified isotope must

. be within the uncertainty specified in the source certification. Validation of system performance was made prior to starting the counting tasks, and upon completion of all counting work for Hatch Unit 2. The counting system performance was acceptable in each case, indicating that the counting system properties did not change during the course of the counting procedure.

Table A-3 shows the counting schedule established for this work. There was no requirement for order of counting since the dosimeter materials still contained sufficient quantities of activation products to allow accurate radio assay. Counting times were more than sufficient to achieve the desired statistical accuracy for gamma emissions of interest in all cases.

Neutrons interact with the constituent nuclei of the dosimeter materials producing radioriuclides in varying amounts depending on total neutron fluence, its energy spectrum, and the nuclear properties of the dosimeter materials. Table A-4 lists the reactions of interest and their resultant radionuclide products for.each element contained in the dosimeters. These are threshold reactions involving an n-p or n-'a interaction.

Finally, Table A-5 presents the primary results of interest for flux and fluence determination.

The specific activity units are in dps/mg, which normalizes the activity to dosimeter mass. The activitie~ are specified for a useful reference date/time, which in this case is the Hatch Unit 2 plant shutdown date and time. This reference date/time was specified as February 6, 2017, at 12:00:00 AM eastern standard time.

A-2

Dosimeter Analysis G4Fe G4Fe Figure A-1 Hatch Unit 2 120° Capsule Packet G4 Fe Dosimeter Wire G4 Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right)

G4Cu G4Cu Figure A-2 Hatch Unit 2 120° Capsule Packet G4 Cu Dosimeter Wire G4 Cu: Prior to Cleaning (left);

and After Cleaning/Coiling (right)

G4Ni G4Ni Figure A-3 Hatch Unit 2 120° Capsule Packet G4 Ni Dosimeter Wire G4 Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right)

A-3

Dosimeter Analysis G5Fe G5Fe Figure A-4 Hatch Unit 2 120° Capsule Packet GS Fe Dosimeter Wire GS Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right)

G5Cu G5Cu Figure A-S Hatch Unit 2 120° Capsule Packet GS Cu Dosimeter Wire GS Cu: Prior to Cleaning (left);

and After Cleaning/Coiling (right)

G5Ni G5Ni Figure A-6 Hatch Unit 2 120° Capsule Packet GS Ni Dosimeter Wire GS Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right)

A-4

Dosimeter Analysis Table A-1 Hatch Unit 2 120°. Capsule Charpy Packet Dosimeter Wire Masses Wire Dosimeter ID Mass (mg)

G4 Fe 1.52.57 G4.Cu 346.21 G4Ni 311.13 G5 Fe 156.67 I

G5Cu 351.92 G5Ni 308.52 Table A-2 Gamma Ray Spectrometer System (GRSS) Specifications System Component Description.and/or Specifications Detector Canberra Model BE3830 Energy Resolution

<1.9 keV FWHM @ 1.33 MeV Detector Efficiency Relative to a 3 inch x 3 inch 33.3% at 1.3 MeV Nal Crystal Amplifier/Multichannel Analyzer Canberra DAS-1000 Computer System Intel i5-4460 CPU at 3.20 GHz, 16 GB Main Memory, 931 GB Hard Disk, 23-inch Monitor, HP LaserJet Printer Software Canberra Apex v 1.4 Table A-3 Counting Schedule for Hatch Unit 2 120° Capsule Dosimeter Materials Dosimeter ID Count Start Date Count Start Time (EST)

Count Duration (Live Time Seconds)

G4.Fe 3/17/2017

10:25:.47 AM 86,400.'

G4Cu 3/20/2017 9:23:41 AM 86,400 G4Ni 3/21/2017 11 :01:35 Arvi 86,400 GS Fe 3/22/2017 3:22:32 PM 86,400 G5Cu 3/23/2017 4:26:28 PM 86,400 G5Ni 3/24/2017 5:16:10 PM 86,400 A-5

Dosimeter Analysis TableA-4 Neutron-Induced Reactions of Interest Dosimeter Material Neutron-Induced Reaction Reaction Product Radionuclide Iron Fe54(n,p)Mn54 Mn54 Copper Cu63(n,a)Co60 cas6 Nickel N i5B( n' p )Co5B Co5s TableA-5 Results of Hatch Unit 2 120° Capsule Radiometric Analysis Activity at

Specific Activity at Activity Dosimeter ID Isotope ID Reference Reference Uncertainty Date/Time 1 (µCi)

Date/Time1 (dps/mg)

(%)

G4 Fe 54Mn 4.82E-01 116.89 2.20 G4Cu soco 2.07E-01 22.12 1.70 G4Ni 5BCo 1.20E+01 1427.06 2.26 GS Fe 54Mn 4.S9E-01 108:40 2.20 GS Cu soco 2.09E-01 21.97 1.70 GS Ni 5BCo 1.16E+01 1391.16 2.26 1 February 6, 2017 at 12:00:00 AM EST is the reference date and time.

A-6

'I

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